0

ASME Conference Presenter Attendance Policy and Archival Proceedings

2018;():V01BT00A001. doi:10.1115/PVP2018-NS1B.
FREE TO VIEW

This online compilation of papers from the ASME 2018 Pressure Vessels and Piping Conference (PVP2018) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Codes and Standards: Probabilistic and Risk-Informed Methods for Structural Integrity Assessment

2018;():V01BT01A001. doi:10.1115/PVP2018-84400.

An approach is outlined for the treatment of stresses in complex three-dimensional components for the purpose of conducting probabilistic creep-fatigue lifetime assessments. For conventional deterministic assessments, the stress state in a plant component is found using thermal and mechanical (elastic) finite element (FE) models. Key inputs are typically steam temperatures and pressures, with the three principal stress components (PSCs) at the assessment location(s) being the outputs. This paper presents an approach which was developed based on application experience with a tube-plate ligament (TPL) component, for which historical data was available. Though both transient as well as steady-state conditions can have large contributions towards the creep-fatigue damage, this work is mainly concerned with the latter. In a probabilistic assessment, the aim of this approach is to replace time intensive FE runs with a predictive model to approximate stresses at various assessment locations. This is achieved by firstly modelling a wide range of typical loading conditions using FE models to obtain the desire stresses. Based on the results from these FE runs, a probability map is produced and input(s)-output(s) functions are fitted (either using a Response Surface Method or Linear Regression). These models are thereafter used to predict stresses as functions of the input parameter(s) directly. This mitigates running an FE model for every probabilistic trial (of which there typically may be more than 104), an approach which would be computationally prohibitive.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A002. doi:10.1115/PVP2018-84767.

In order to address the risks associated with the operation of ageing pressure boundary components, many assessments incorporate probabilistic analysis tools for alleviating excessive conservatism of deterministic methodologies. In general, deterministic techniques utilize conservative bounding values for all critical parameters. Recently, various Probabilistic Fracture Mechanics (PFM) codes have been employed to identify governing parameters which could affect licensing basis margins of pressure retaining components. Moreover, these codes are used to calculate a probability of failure in order to estimate potential risks under operating and design loading conditions for the pressure retaining components experiencing plausible and active degradation mechanisms.

Probabilistic approaches typically invoke the Monte-Carlo (MC) method where a set of critical input variables are randomly distributed and inserted in deterministic computer models. Estimates of results from probabilistic assessments are then compared against various assessment criteria.

During the PVP-2016 conference, we investigated the assumption of normality of the Monte Carlo results utilizing a non-linear system function. In this paper, we extend the study by employing non-normal input distributions and investigating the effects of sampling region on the system function.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A003. doi:10.1115/PVP2018-84964.

In Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWRs) for structural integrity assessment of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock transients. By reflecting the latest knowledge and findings, the PASCAL code has been continuously improved. In this paper, the development of PASCAL Version 4 (hereafter, PASCAL4) is described. Several analysis functions incorporated into PASCAL4 for evaluating the failure frequency of RPVs are introduced, for example, the evaluation function of confidence level of failure frequency considering epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions and KI calculation methods considering complicated stress distributions, and the recent Japanese irradiation embrittlement prediction method. Finally, using PASCAL4, a PFM analysis example for a Japanese model RPV is presented.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A004. doi:10.1115/PVP2018-85086.

Requirements for pressure-temperature limits to protect against rupture of CANDU nuclear reactor Zr-Nb pressure tubes are provided in the Canadian Standards Association (CSA) Standard N285.8. The requirements are based on a stability evaluation of a postulated axial through-wall flaw for all ASME Service Level A, B, C and D loadings. The flaw stability evaluation is strongly dependent on the fracture toughness of the Zr-Nb pressure tube material. The fracture toughness of Zr-Nb pressure tubes is decreasing with operating hours. The decrease in fracture toughness as well as compounding conservatisms based on using bounding values make deterministic evaluations more challenging. The CSA Standard N285.8 permits probabilistic evaluations of fracture protection, but does not provide acceptance criteria. Proposed acceptance criteria that meet the intent of the design basis for Zr-Nb pressure tubes have been developed. The proposed acceptance criteria consist of a proposed maximum allowable conditional probability of pressure tube rupture for the entire reactor core, as well as a proposed maximum allowable conditional probability of rupture of a single pressure tube. The paper provides a description of the technical basis for the proposed acceptance criteria for probabilistic evaluations of fracture protection.

Commentary by Dr. Valentin Fuster

Codes and Standards: Quality Assurance

2018;():V01BT01A005. doi:10.1115/PVP2018-84109.

Welds are inspected by various techniques which include visual examination, surface examination and volumetric examination.

While the above techniques would qualify a weld to workmanship criteria, they would not necessarily be indicative of weld properties.

Preparation and qualification of welding procedures and testing of production welds are indicative that the weldment would probably provide a safe and satisfactory service life.

However, weldments have to operate at their design conditions which may include high temperatures and ASME Codes do not necessarily stipulate tests for verification of high temperature properties. In addition, defective welds are often repaired by removing the originally deposited weld metal and re-welding. The effects of double heat input are not necessarily evaluated.

In this paper, an insight is provided into the factors which provide assurance that weldments will perform satisfactorily in service and the combination of non-destructive evaluation methods which would enable effective detection of imperfections.

Non-destructive volumetric examination method for welds has traditionally be radiography. With the advent of automated data acquisition methods in Ultrasonics, like Time of Flight Diffraction and Phased Array Ultrasonic Testing, these methods are rapidly replacing radiographic methods for weld inspection.

Ultrasonic acceptance criteria in ASME Section VIII Div. 1, ASME Section VIII Div. 2 and ASME Section IX do not include evaluation of porosity as ultrasonic methods do not easily detect porosity.

The result of all this is that today we are accepting welders qualified using Ultrasonic examination as per ASME Section IX but on the job there is still the option of inspecting the weld using Radiography in which, excessive porosity can be a cause for weld repair.

Considering this and various other criteria, a comprehensive weld evaluation methodology is proposed taking advantage of the strengths of each inspection technique while welding technology used would ensure that welds have required properties at service temperatures.

A proposal is also made to improve the detectability of imperfections using modifications of existing Ultrasonic A-scan Techniques.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A006. doi:10.1115/PVP2018-85160.

Inter-laboratory proficiency testing of NDT laboratories acc. to ISO 17043 is an effective tool to select laboratories competent to perform the NDT tasks according to the customer requirements in desired quality especially. The goal of this article is to show that inter-laboratory proficiency testing is also a relevant tool for assessing the continuous improvement principle acc. to ISO 9001. It is demonstrated how proficiency testing may help the NDT labs to fully understand the NDT process, get feedback and self-reflection on performing NDT tasks, monitor the internal efficiency and improve the laboratory through a continuous evolution of NDT process.

Topics: Testing
Commentary by Dr. Valentin Fuster
2018;():V01BT01A007. doi:10.1115/PVP2018-85161.

ASME (American Society of Mechanical Engineers) Nondestructive Examination (NDE) and Quality Control (QC) Central Qualification and Certification Program (ANDE-1 Standard) is a new independent qualification system for NDT and QC personnel. It is the first to come with a requirement of performance-based experience of Nondestructive Testing (NDT) personnel on the basis of practical demonstration to perform all required activities rather than length of experience. It also adds unflawed samples to the test sets in order to simulate real situation in industrial practice, where decisions about the acceptance is necessary. This article analyzes the ANDE-1 Standard and compares it with other relevant qualification standards used in the European Union (EU).

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in ASME Codes and Standards

2018;():V01BT01A008. doi:10.1115/PVP2018-84031.

ASME Nuclear Codes and Standards are used worldwide in the construction, inspection, and repair of commercial nuclear power plants. As the industry looks to the future of nuclear power and some of the new plant designs under development, there will be some significant departures from the current light water reactor (LWR) technology. Some examples are gas-cooled and liquid metal-cooled high temperature reactors (HTRs), small modular reactors (SMRs), and fusion energy devices that are currently under development. Many of these designs will have different safety challenges from the current LWR fleet.

Variations of the current LWR technology are also expected to remain in use for the foreseeable future. Worldwide, many LWRs are planned or are already under construction. However, technology for construction of these plants has advanced considerably since most of the current construction codes were written. As a result, many modern design and fabrication methods available today, which provide both safety and economic benefits, cannot be fully utilized since they are not addressed by Code rules.

For ASME Nuclear Codes and Standards to maintain and enhance their position as the worldwide leader in the nuclear power industry, they will need to be modernized to address these items. Accordingly, the ASME Nuclear Codes and Standards organizations have initiated the “2025 Nuclear Code” initiative. The purpose of this initiative is to modernize all aspects of ASME’s Nuclear Codes and Standards to adopt new technologies in plant design, construction, and life cycle management. Examples include modernized finite element analysis and fatigue rules, and incorporation of probabilistic and risk-informed methodology. This paper will present the vision for the 2025 ASME Nuclear Codes and Standards and will discuss some of the key elements that are being considered.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A009. doi:10.1115/PVP2018-84076.

ASME B31.3 Appendix L provides the pressure and temperature ratings of forged aluminum flanges. The flanges are from NPS 1/2 to NPS 24 in three rating Classes 150, 300 and 600 with two grades of aluminum alloys: ASTM B247 3003-H112 and 6061-T6. However, B31.3 does not provide any technical information on the basis of the pressure and temperature ratings. A review of the historical development of ASME B16.5 indicated that the aluminum flanges had the same technical basis for pressure and temperature ratings as the ferrous alloy flanges in ASME B16.5. The 1960 Addenda of the 1957 Edition B16.5 included both aluminum flanges and ferrous alloy flanges. A new Code Case 2905 has been recently approved to allow B31.3 Appendix L aluminum flanges in fabricating Section VIII Division 1 pressure vessels as B16.5 flanges on the basis that both flange specifications have the same safety margin. In this paper, the technical basis of the pressure and temperature rating of aluminum flanges is revisited. Based on the same principle, the pressure ratings are extended to Class 900 and Class 1500 for the two aluminum alloys using the same analysis. Since ASTM B247 5083-H112 is another common grade of aluminum forging alloy, the pressure and temperature ratings are proposed for 5083-H112.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A010. doi:10.1115/PVP2018-84092.

Code Case N-513 provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. The application of the Code Case is restricted to moderate energy, Class 2 and 3 systems, so that safety issues regarding short-term, degraded system operation are minimized. The first version of the Code Case was published in 1997. Since then, there have been four revisions to augment and clarify the evaluation requirements and acceptance criteria of the Code Case that have been published by ASME. The technical bases for the original version of the Code Case and the four revisions have been previously published [1, 2, and 3].

There is currently work underway to incorporate additional changes to the Code Case and this paper provides the technical basis for the changes proposed in a fifth revision. These changes include clarification for buried piping, investigation of various radii used in the Code Case, removal of the 0.1 limit on the flexibility characteristic for elbow flaw evaluation, and an update of the stress intensity factor parameters for circumferential through-wall flaws. In addition, a new flaw evaluation procedure is given for through-wall flaws in gate valve body ends. This procedure evaluates flaws in the end of the valve body as if in straight pipe. These changes and their technical bases are described in this paper. Clarifications and changes deemed editorial are not documented in this paper.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A011. doi:10.1115/PVP2018-84100.

High temperature nuclear reactors operating in the creep regime are designed to withstand numerous cyclic events. Current ASME code rules provide two basic paths for evaluating creep fatigue and ratcheting under these conditions; one based on full inelastic analysis intended to provide a representative stress and strain history and the other based on elastic material models with adjustments of varying complexity to account for inelastic stress and strain redistribution. More recent developments have used elastic-perfectly plastic analysis to bound the effects of cyclic service. However, these methods still rely on the separate evaluation of fatigue and creep damage utilizing a damage interaction diagram. There is a procedure under current development that uses creep-fatigue data from key feature test articles directly without the use of the damage interaction diagram. However, it requires a reasonable representation of the strain range in a structure as an input. This work develops a simplified procedure based on elastic perfectly-plasticity analysis that can be used to represent the strain range in a structure in the steady state under cyclic loading conditions.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A012. doi:10.1115/PVP2018-84101.

Cladding structural components with a corrosion resistant material may greatly extend the design life of molten salt reactor concepts. A complete design methodology for such cladded, high temperature nuclear components will require addressing many issues: fabrication, corrosion resistance, metallurgical interaction, and the mechanical interaction of the clad and base materials under load. This work focuses on the final issue: the mechanical interaction of the base and clad under creep-fatigue conditions. Depending on the relative mechanical properties of the two materials the clad may substantially influence the long-term cyclic response of the structural system or its effect might be negligible. To quantify the effect of different clad material properties we develop an efficient method for simulating pressurized cladded components in the limiting case where the section of interest is far from structural discontinuities. Using this method we evaluate the mechanics of the clad/base system and identify different regimes of mechanical response. The focus is on situations relevant to high temperature nuclear components: thermal-cyclic Bree-type problems and similar axisymmetric structures. The insights gained from these structural studies will form the basis for developing design rules for high-temperature, nuclear, cladded components.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A013. doi:10.1115/PVP2018-84102.

The distinction between a ratcheting and non-ratcheting response is critical for many high temperature design methods. Non-ratcheting is generally considered safe — deformation remain bounded over the lifetime of the component — while ratcheting is undesirable. As a particular example, the elastic perfectly-plastic (EPP) design methods described in recent ASME Section III, Division 5 code cases require a designer to distinguish ratcheting from non-ratcheting for finite element analyses using a relatively simple, elastic perfectly-plastic constitutive response. However, it can be quite difficult to distinguish these two deformation regimes using finite element (FE) analysis particularly in the case where the actual ratcheting strain is small. In practice FE analysis of structures that are analytically in either the plastic shakedown or ratcheting regimes will result in small, cycle-to-cycle accumulated strains characteristic of ratcheting. Distinguishing false ratcheting — the structure is actually in the plastic shakedown regime — from true ratcheting can be challenging. We describe the characteristics of nonlinear FE analysis that cause these false ratcheting strains and describe practical methods for distinguishing a ratcheting from a non-ratcheting response.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A014. doi:10.1115/PVP2018-84103.

Design approaches using elastic perfectly-plastic (EPP) analysis have recently been approved as Code Cases for the Section III, Division 5 design of high-temperature nuclear reactor components made from austenitic stainless steel. These methods bound the ratcheting strain and creep-fatigue damage accumulated over the life of a component with a simplified, elastic-perfectly plastic analysis using a special pseudo-yield stress — often not equal to the true material yield stress. The austenitic materials specified in the existing Code cases are cyclic-hardening for all allowable operating temperatures. However, other Section III, Division 5 materials, such as Grade 91 steel, are cyclic softening at expected advanced reactor operating temperatures. This work describes the extension of EPP methods to cyclic softening materials through the use of a postulated saturated material state and softening factors to be applied to the pseudo yield stress. We demonstrate the conservatism of the modified EPP method against a series of inelastic simulations of two bar tests, using a constitutive model that captures work and cyclic softening.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A015. doi:10.1115/PVP2018-84104.

Grade 91 steel has been called out for use in advanced reactor intermediate heat exchangers and other components. The material has good high temperature creep resistance and thermal properties but has a complex microstructure that manifests as cyclic softening, work softening, and tension/compression asymmetry in its engineering mechanical response. We describe a unified viscoplastic model for the deformation of Grade 91 for an expected operating temperature range spanning from room temperature to approximately 650°C. The model transitions from a rate independent response at low temperatures and high strain rates to a rate dependent, unified viscoplastic response at high temperatures and low creep strain rates. The model captures work and cyclic softening in the material through combined isotropic-kinematic hardening and captures observed tension/compression asymmetry and related anomalous ratcheting effects through a non-J2 flow term. A particular focus of the model is on capturing the average response of Grade 91 as determined from a wide collection of experimental data at many different temperatures, rather than the response of a single set of experiments at a particular temperature. The final model is suitable for the engineering design of nuclear components via inelastic analysis using the ASME Section III, Division 5 procedures.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A016. doi:10.1115/PVP2018-84105.

Section III, Division 5 of the ASME Boiler and Pressure Vessel Code provides two broad paths for the design of high temperature, safety-critical nuclear components: design by elastic analysis and design by inelastic analysis. The design by elastic analysis approach, as the name suggests, uses a linear elastic stress analysis of the component and applies design rules designed to bound response of the actual structure, which will undergo both creep and plasticity. Currently, the Code allows the use of the elastic approach for all operating temperatures up to the maximum use temperatures in the Code. The bounds used in the elastic approach assume an uncoupled material response combining rate dependent creep with rate independent plasticity. However, at elevated temperatures creep and plasticity are coupled, rate dependent mechanisms and so the elastic analysis rules may become non-conservative. We present several examples of potential non-conservatism in the elastic analysis rules at high operating temperatures. Then we describe a systematic method for determining a temperature cutoff describing the transition from non-unified, rate independent plasticity material response to a rate dependent, unified plastic response. Logically, this transition temperature sets the upper bound for the allowable, conservative use of the design by elastic analysis approach and so we propose these temperatures, determined for all the Section III, Division 5 Class A materials, as Code limits for the applicability of the elastic approach.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A017. doi:10.1115/PVP2018-84106.

The Bree solution to the problem of a ratcheting cylinder under constant pressure and cyclic thermal load is a fundamental result in nuclear engineering and widely used as the technical basis for the ASME Boiler and Pressure Vessel Code and other design methods. However, because the loading conditions in the Bree problem are difficult to achieve experimentally there have been relatively few works experimentally examining the problem and extending it to other relevant design situations, for example cladded components. In contrast, 2-bar problems are widely studied experimentally and are relatively easy to setup. These 2-bar problems are thought to be representative of Bree-type geometries, but a formal connection has not been demonstrated. This work formally establishes the connection between the Bree cylinder and an n-bar problem — a coupled bar experiment with, in general, more than two bars linked in parallel. The connection suggests that n-bar experiments using a fairly limited number of bars might be an experimentally-accessible setup that better represents ratcheting phenomenon in actual nuclear pressurized components. Such experiments could test surrogate cladded or multi-material components by using bars of different materials. Finally, this work suggests control schemes that yield optimally efficient n-bar experiments — experiments that best replicates a Bree cylinder with a limited number of bars.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A018. doi:10.1115/PVP2018-84793.

The use of surveillance specimens was identified in a roadmap (Sims, 2010, ASME HTGR Code Development Roadmap, STP-NU-045, American Society of Mechanical Engineers, New York, NY) developed for the U.S. Nuclear Regulatory Commission for non-light water reactors that are intended to have very long design lives on the order of 500,000 hours. Creep-rupture, long term strain accumulation, creep-fatigue damage, and their interaction with environmental effects require significant extrapolation from shorter term tests. Surveillance specimens provide a means for confirmation of these extrapolations in-situ. The effects of cyclic service on creep-fatigue damage would be particularly challenging based on conventional mechanical testing. Proposed herein is a passive cyclic testing methodology that lends itself well to in-situ surveillance applications. The concept is based on utilizing the difference in the thermal expansion coefficient of candidate materials to generate cyclic loading based on operational thermal transients. Prototypical specimen designs are proposed and their in-situ response to representative plant operation are evaluated.

Topics: Surveillance , Water
Commentary by Dr. Valentin Fuster
2018;():V01BT01A019. doi:10.1115/PVP2018-84823.

Very few materials have all their properties behave the same in all orientations of the material. The term for this behavior is anisotropy or anisotropic properties. The properties that are affected and the degree of anisotropy they exhibit, depend on the material family, the alloy, and the processing of the material. This paper will only discuss metals, and limited to those metal specifications adopted in the ASME Boiler & Pressure Vessel Code (B&PVC) Section II, Parts A and B.

The anisotropic properties of plate have been well recognized explicitly and implicitly in the B&PVC for a very long time. At issue, and the specific focus of this paper, is another wrought product form: bars. Bar, specifically round bar — also called rod, is a very useful starting material to manufacture pressure parts. The bar can be stocked at the largest feasible diameter and length, and then cost-effectively machined to whatever diameter and length is desired for the part. Due to the anisotropic properties of bar, there is a need for understanding and setting limitations for the use of such material in the B&PVC. This paper will explore the past history, current status and future directions of the Code requirements for bar.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Chinese Codes and Standards

2018;():V01BT01A020. doi:10.1115/PVP2018-84048.

Buckling behavior of tanks with a conical roof under harmonic settlement has been researched in this paper. A real tank in engineering is taken into account and the harmonic settlement is applied to the bottom of the tank to simulate its buckling behavior. Results show that the tank wall will be subjected to a deformation mutation when the settlement reaches a critical value. It means that compared to the conical roof, the tank wall is more vulnerable to buckling. Because of the complexity added by the grid of rafters and rings on the roof, two different simplified models are presented. For the first model, the tank’s roof is modified to an equivalent thickness based on the smeared method; for the other one, the roof is completely eliminated and its influence is represented by simply supported boundary conditions at the top of the tank wall. Analysis shows that the tank model without a roof can’t reflect buckling behavior of the real tank in engineering very well. While the model with an equivalent thickness roof can avoid this deficiency and achieve high efficiency and accuracy. It’s recommended to be applied to buckling analysis of tanks under settlement. Based on that, effects of wave numbers on the critical settlement for the three models are researched and compared. Result shows that the simplified tank model with an equivalent thickness roof presented in this paper is efficient and useful for buckling analysis of tanks with a conical roof under settlement.

Topics: Buckling , Roofs
Commentary by Dr. Valentin Fuster
2018;():V01BT01A021. doi:10.1115/PVP2018-84059.

As of the end of 2015, more than 20 transportable pressure vessel leakage accidents were put out every day by fire forces in China. Accidents put out by fire forces are primarily handled in accordance with codes and standards. This paper investigates and analyzes the codes and standards systems for pressure vessel and fire emergency rescue (FER) in China. Existing codes and standards for pressure vessel leakage do not include FER, which results in the firefighters being encumbered when addressing these accidents. Treatment methods for hazardous chemicals are included in the FER codes and standards, but they are not aimed at pressure vessels. One of the goals of the Plan for Furthering the Standardization Reforms released by the State Council in 2015 was to integrate and reduce the number of codes and standards. It is impossible to establish new FER standards for pressure vessels. To address this issue, this paper presents a new concept of pressure vessels standard for FER. This new concept features two aspects: (1) China’s Standardization Committee on Boilers and Pressure Vessels should absorb some fire practitioners, in particular fire scientific researchers and firefighters; and (2) the pressure vessel FER technologies, methods, and structural components should be added to the pressure vessel codes and standards.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A022. doi:10.1115/PVP2018-84171.

A cracking incident of a 304 stainless steel elbow serving in the synthesis gas purification device occurred during running. In order to get an understanding of the failure mechanism, a failure analysis was performed on the cracked elbow in this paper. The chemical composition, mechanical properties of strength, toughness and hardness, hydrogen content were identified and determined. The metallographical structure was observed and analyzed by optical microscope (OM) and X-Ray Diffraction (XRD), while the fracture morphology was observed by scanning electron microscope (SEM). The results showed that the chemical composition of the cracked elbow meet the requirements for China standard, while comparing with GB/T 14976-2012 standards, the strength and elongation of the leaked elbow are higher and lower respectively, and the hardness of the leaked elbow was higher than quality certificate documents that of HB ⩽ 187. Large quantities of martensite and δ-ferrite were observed in elbow, which indicated that the elbow was not well solid solution heat treated required by specification (1050°C,30min). The fracture morphology presents typical brittle fracture. The hydrogen content of cracked elbow was significant higher than that of other 304 stainless steel elbow serving in the environment without hydrogen. It is acknowledged that martensite showed higher sensitivity of hydrogen embrittlement compared with austenite. Furthermore, the operating temperature of cracked elbow was in the range of high hydrogen embrittlement sensitivity. Depending on the metallographical structure, strength, service environment, hydrogen content and fracture morphology, it can be concluded that hydrogen induced delayed cracking was the dominant mechanism of the failure.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A023. doi:10.1115/PVP2018-84176.

In recent years, with the gradual expansion of units scale for energy industries such as oil refining, chemical and natural gas, etc. in China, pressure vessels are developing toward the large and heavy direction with the maximum wall thickness reaching hundreds of millimeters and weight exceeding one thousand tons. It not only causes huge material consumption, difficulty in manufacture and processing (even beyond current capability), but probably also lead to new failure modes and mechanisms, and significant safety risks thereby. Therefore, the lightweight design and manufacture under the premise of inherent safety has become an urgent need to break through the bottleneck of manufacturing capability and realize the material and energy conservation. In this paper, the progress of lightweight design and manufacturing technology of heavy-duty pressure vessels in recent years in China is introduced from the aspects of adjustment of material allowable strength, application of high strength steel and matching of strength and toughness, cold stretch of austenitic stainless steel, structure optimization design with multi-parameter coupling, and application of composite materials. Application cases in the construction of pressure equipments such as large hydrogenating reactors, low-temperature ethylene spherical tanks, austenitic stainless steel cryogenic vessels, butyl octanol heat exchangers, etc. are also included. Finally, some suggestions on the future research are proposed including composite pressure vessels in cryogenic environment, materials genome and additive manufacturing technologies for pressure vessels.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A024. doi:10.1115/PVP2018-84201.

Austenitic stainless steel butt joints are widely used in the pressure piping system, and the quality of welded joints directly affect the safety of pressure special equipment. In this paper, phased array ultrasonic testing technology is used to study the feasibility of 4mm∼10mm wall thickness workpiece. Through the software CIVA (Developed by The French Alternative Energies and Atomic Energy Commission (CEA)) simulat to determine the parameters of the detection system, and it tests the 18 groups of 4mm∼10mm series simulation samples by PAUT (Phased Array Ultrasonic Testing).Through comparison with Radiographic testing, PAUT for the girth weld can be effectively for the 4mm∼10mm Austenitic Stainless Steel Pipeline.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A025. doi:10.1115/PVP2018-84262.

A failed carbon steel elbow from a natural gas gathering pipeline in a gas field in Northeast China was investigated by macroscopic and microscopic examinations, chemical composition analysis, metallographic examination, and numerical simulation methods. The investigation results show that the intrados of the elbow was subject to slight general corrosion, while the extrados suffered from severe localized corrosion. The damage of the elbow resulted from an erosion-corrosion in the natural gas containing a few amount of corrosive impurities, liquid water, and solid particles. The impurities in the natural gas, specifically CO2 and chlorides, would be dissolved into water droplets in the natural gas. These corrosive droplets reacted with the pipe metal, resulting in typical CO2 corrosion of carbon steel pipe. Furthermore, the droplets and solid particles in the gas would destroy the protectiveness of the corrosion product film on the intrados by mechanical erosion, finally leading to the deterioration of the local environment and then the acceleration of corrosion failure. For controlling corrosion, some measures should be given. However, considering the difficulty of the increase in the curvature radius or the internal diameter of the pipeline, increasing wall thickness of the elbow pipe was a relatively feasible measure to mitigate the erosion-corrosion of the pipe.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A026. doi:10.1115/PVP2018-84265.

This work is to address the creep analysis for components at elevated temperatures based on isochronous stress-strain curve and the elastic-perfectly plastic material model through numerical analyses. Numerical results presented that the creep deformation is very sensitive to the target inelastic strain chosen for analysis. A small inelastic strain, corresponding to a small yield stress, can cause very conservative result for the case studied. Moreover, the target inelastic strain, corresponding to the minimum inelastic strain along with the given path, is different from each other for various internal pressures.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A027. doi:10.1115/PVP2018-84275.

Due to the Small diameter and thin walled tube docking girth joint ultrasonic testing has the problems of: dead zone is often greater than the wall thickness; large curvature of the pipeline causes ultrasonic scattering, greatly reduces the sensitivity; pipe weld root’s is not welded and so on. In this paper, the prepared small-bore and sheet steel tube docking girth joint ultrasonic testing probe has sufficient sensitivity, is less than 2.5mm of initial pulse width, short front (≤ 5mm), high resolution, less clutter, less surface wave components, At the same time, the test block was designed and used for the testing of probe’s related technical indicators, the test block can meet the focus probe’s performance and scanning sensitivity calibration. Combined with the field detection and compared with the radiographic testing, the ultrasonic testing obtain good result. It is suitable for the ultrasonic quantitative testing of the joint ring weld of thin wall pipe.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A028. doi:10.1115/PVP2018-84331.

2¼Cr-1Mo-¼V steel is used for manufacturing and producing hydrogenation reactors. The coarse-grained heat affected zone (CGHAZ) of welded joints using 2¼Cr-1Mo-¼V steel have a higher reheat cracking susceptibility than the conventional 2¼Cr-1Mo steel. But there is no simple and effective method evaluating reheat cracking susceptibility of CGHAZ in 2¼Cr-1Mo-¼V steel welded joints at present. Based on Weld C-Ring Test and Gleeble® Thermo-Mechanical Test, a new Notched C-Ring Test to evaluate reheat cracking susceptibility of CGHAZ in 2¼Cr-1Mo-¼V steel was studied in this paper. According to the fitting relationship between RoA and critical stress, a criteria was proposed preliminarily to determine the reheat cracking susceptibility of 2¼Cr-1Mo-¼V steel CGHAZ using “not susceptible” and “slightly susceptible”.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A029. doi:10.1115/PVP2018-84335.

Thermal load is one of the most important design conditions that should be considered carefully in engineering practice. Most inner-pressure vessels suffer thermo-mechanical ratcheting or unacceptable plastic deformation under cyclic thermal stress produced by inside heat source and pressure-induced primary stress. However, thermal load is also a crucial factor for external-pressure vessels where the failure model of buckling should not be ignored. The effect of thermal load on buckling is not only thermal stress itself but also shape distortion due to thermal load. In some cases, the latter is more important. In this paper, an external-pressure thin-walled ellipsoidal head with heating jackets has been studied. The temperature of this structure is uniformly distributed along the thickness direction but changes alternately between hot and cold along the meridional direction, which will have a significant effect on buckling behavior of this typical structure. Buckling load is sensitive to initial defect and small deformation. Several comparative calculations based on nonlinear buckling analysis have been conducted and some laws are established. Finally, some useful conclusions and suggestions are proposed for engineering design.

Topics: Stress , Buckling , Shapes
Commentary by Dr. Valentin Fuster
2018;():V01BT01A030. doi:10.1115/PVP2018-84378.

Quenched and tempered high strength steel 07MnMoVR with a better combination of mechanical properties and low susceptibility to welding crack has attracted attention for applications in engineering fields. Exposure to fire will subject steel to thermally induced environmental conditions that may alter the material’s properties. The residual strength after a fire event is important to assess the extent of the fire damage and the potential reusability of the vessel. This paper presents the details of an experimental investigation on the post-fire mechanical properties of 07MnMoVR steel. Uniaxial tension tests and Charpy impact tests were performed on coupons exposed to elevated temperatures varying from 550°C to 850°C for half an hour to 8 hours and then naturally cooled in air or cooled by water. The post-fire stress-strain curves, strength, ductility and impact toughness of 07MnMoVR steel are discussed. The results show that the yield plateau in post-fire stress-strain curves disappears when the exposure temperature is higher than 700°C. The residual yield strength and ultimate strength decrease firstly and increase afterward with increasing exposure temperature. The influences of duration time on the residual strength are considerable for exposure at 650°C. The post-fire impact toughness of 07MnMoVR steel at −20°C degrades drastically with increasing duration time when the exposure temperature reaches 700°C. The effects of cooling methods on strength and toughness become significant when the exposure temperature exceeds 750°C. The critical tempetature for the mechanical properties deterioration is 650°C. This study can provide basis data and guidelines for the fitness for service assessment of 07MnMoVR steel suffered from fire accident.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A031. doi:10.1115/PVP2018-84437.

In this paper, the mechanical properties and microstructural changes of 2.25Cr1Mo0.25V steel under different heat treatment and welding process were investigated. The heat treatment of steel during practical processing is taken as a reference. Different heat treatment time are used to obtain samples with different condition. Automatic submerged arc welding was used to obtain welding sample. The mechanical properties of different samples are obtained by tensile test; the evolution of microstructure and precipitates of different sample with heat treatment and welding was studied on scanning electron microscopy. The experimental results show that with the increase of heat treatment time, the strength of the samples decreases and the plasticity remains nearly constant. Heat treatment also affects the precipitation of carbides; the longer the heat treatment time is, the more precipitates are. Compared with the base metal, the welding metal sample has higher strength. The amount of precipitates in welding metal is much larger than it in base metal. The research on precipitation shows that there are different kinds of precipitates which have different morphologies in welding metal.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A032. doi:10.1115/PVP2018-84451.

Pressure Strengthening (PS) is a technology widely used to increase the allowable stress and thus reduce the weight of Austenitic Stainless Steel (ASS) cryogenic vessels. Pre-strain and low temperature are two important factors enhancing the strength. However, standards such as EN 13458-2, ASME VIII-I, AS 1210 and ISO 21009-1, state that the strengthening is based on work hardening of ASS, while the effect of low temperature was ignored or not mentioned. Therefore, in this work, the influence of low temperature on cryogenic mechanical properties of S30408 stainless steel and its welded joints were studied firstly, then a numerical analysis on the influence of pressure strengthening of a large transportable cryogenic storage tank was conducted. Finally, after the technical significance of PS was discussed, a new understanding was proposed. Conclusions come that: (a) At room temperature, the strength of ASS can be increased by pre-strain; while at −196 °C, the strengthening of low temperature plays a leading role (2/3 or more) for strength increment; (b) The purpose of PS technique is to stabilize the dimensions of the internal vessel. It is unnecessary to consider insufficient strengthening or over strengthening at different location because the strength enhancement is mainly due to low temperatures.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A033. doi:10.1115/PVP2018-84455.

The vertical vessels supported by skirt are sensitive to dynamic loads, that might oscillate intensely under earthquake and wind loads. Therefore, calculating the dynamic loads and response accurately, especially the damping ratio, influences the amplitude of vessels vibration greatly and is vital for safe operation of the vertical vessels. In Chinese design code, the damping ratio is 0.01 for steel structure, and in American code, it consists of structural damping βs and aerodynamic damping βa, while in European code, the additional damping βd is also considered. With full-scale field tests, the damping ratios of vertical vessels, through a thorough data collection, is recommended as 0.002∼0.005. The damping ratio is 2/3 less than the design value 0.01 in empty condition, that makes the vessel in a crucial state. A new design concept is proposed that the damping ratio of vertical vessels should be determined based on various conditions. Fatigue checking is required in empty condition besides designing by code in operating condition. In addition, the allowable placement time [Tp] is calculated in empty condition. The actual placement time Tp should satisfy Tp < [Tp] and the damper should be installed on the vertical vessel to provide additional damping ratio in the case of Tp > [Tp]. An experiment was carried out to illustrate the new design concept that the damping ratio was enlarged by installing dampers. The damping ratio with other research could provide fundamental parameters for the amendment of the design code.

Topics: Damping , Design , Vessels
Commentary by Dr. Valentin Fuster
2018;():V01BT01A034. doi:10.1115/PVP2018-84457.

The pressures relief devices (PRDs) are widely used on long tube trailers, and can automatically open in a fire accident to relieve pressure when the pressure exceeds the set value to keep the cylinder safe from explosion. However, there is a big difference in the structure selection of PRDs and the calculation methods of the effective discharge area between different standards such as GB/T33215, API521 and CGA S-1.1. The overall fire test and local fire test of large capacity steel seamless cylinder were carried out to obtain the response behavior of different PRDs with different discharge areas, and the change law of temperature and pressure during the pressure relief process. Results showed that the requirements of safety release of cylinders in the fire environment were satisfied by the effective discharge area calculated by API521, CGA S-1.1 and GB/T33215, and the inner temperature and pressure were greatly affected by the different discharge orifices. In the same cases that the PRD were isolated from fire by steel plate, the single rupture disk device has a faster response than the combination rupture disk/fusible-plug devices.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A035. doi:10.1115/PVP2018-84460.

In new design of heat exchangers with high turbulence intensity and low pressure drop in the shell side, parameters or structures beyond the standards may be adopted and failure risk due to fluid induced vibration can increase. As pitch ratio is related with some significant parameters, the range of pitch ratio determines whether the FIV calculation could be established. In this research, FIV of heat exchangers with pitch ratio beyond the standards was investigated on mechanisms such as vortex shedding, turbulent buffeting and fluid elastic instability. Both theoretical and numerical methods have been applied during the study to get parameters beyond the pitch ratio range. Parameters related to pitch ratio such as additional mass coefficient Cm and damping ratio ζ were extended, and then verified by experiments. Calculating cases on several conditions were accomplished as well as comparation with literatures. The results and methods could be guidelines for the design of new type of heat exchangers and references in the amendment of relevant standards.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A036. doi:10.1115/PVP2018-84497.

Butt thermal-fusion joint is one of the main joints for polyolefin pipes. It could lead to different failure modes caused by factors such as different environments and operations, which will makes the joints become one of the most dangerous potential risks in polyolefin pipe system. As different defects could lead to different failure modes, defect recognition and classification are considered to be the critical issue for ultrasonic inspection. Based on the previous researches about defect recognition, simulation and experiments on artificial specimens in ultrasonic inspection, the imaging mechanism and identifying method were investigated for different defects, and the corresponding relations between joint defect characteristics and ultrasonic images were proposed. Furthermore, a defect recognition method was established. Many defect inspection, identification and dissection experiments were conducted on unknown defects, and the proposed method of automatic defect recognition was verified.

Topics: Inspection
Commentary by Dr. Valentin Fuster
2018;():V01BT01A037. doi:10.1115/PVP2018-84498.

This paper is concerned with the low-cycled fatigue life of S30408 austenitic stainless steel at 77 K. Strain-controlled low-cycled fatigue tests were performed in a liquid-nitrogen bath covering a strain-amplitude range of 0.4%–1.0%. The role of the reduced temperature is examined during the low-cycled fatigue tests by comparing the fatigue performance to the one at ambient temperature that was obtained in our previous work. It is found that the cryogenic low-cycled fatigue life is significantly improved by a factor of 5–10 in the low strain-amplitude range of 0.4%–0.5%, resulting from the pronounced hardening effect due to the low temperature. However, the cryogenic improvement gradually reduces with the increasing strain-amplitude. At 77 K, the cyclic stress amplitude increases rapidly at the beginning of the fatigue test, and no cyclic softening is found due to the cryogenically constrained movement of the dislocations. The fatigue hysteresis loops and fatigue stress-strain curves shows that the cyclic plastic strain at cryogenic temperature accounts for a limited proportion in the total cyclic strain, and the damage may occurs explosively at the beginning of the cyclic load at 77 K.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A038. doi:10.1115/PVP2018-84507.

The operating environment of polyethylene (PE) gas pipeline is complex and changeable. In the process of pipeline design, construction and operation, there are many fuzzy factors inducing pipeline accidents. In the paper, the fuzzy factors causing pipeline accidents were studied and analyzed, and the risk evaluation factors of PE gas pipeline were determined. According to the theory of fuzzy comprehensive evaluation, the mathematical model of fuzzy comprehensive evaluation for PE pipeline was established. In order to pursue as much accurate results of assessment as possible, the fuzzy analytic hierarchy process was used to determine the weight of each index. The safety evaluation grade of the pipeline was given by using the risk matrix method. And then the evaluation system software of PE gas pipeline risk assessment was developed. The software has been used in the risk inspection of PE gas pipeline, which provides the basis for the risk management of the pipeline.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A039. doi:10.1115/PVP2018-84527.

A company’s isocyanate and polyether MMDI distillation column is a Dividing wall column (called DWC). During the trial operation, the dividing wall is distorted so that the distillation can not work on. Finite element method was used in this subject, considering the dividing wall with no flaw, unevenness and with initial stress, studying the deformation during the temperature loading, unloading and re-loading respectively. According to the results, main factors for the deformation of dividing wall were analyzed and possible solutions were proposed eventually.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A040. doi:10.1115/PVP2018-84548.

Ultrasonic guided wave inspection technology has been widely for long distance pipeline inspection; however, the pipe elbow’s discontinuous structure and the dispersion of L-type wave are restricting the application of this technology. This paper proposes a method of L(0,2) mode guided wave excitation based on magnetostrictive effect and explores the optimization of the magnetization sensor arrangement. Test results shows that the proposed method can detect many types of defects in the pipe elbow. This paper encourages the use of L(0,2) mode guided wave excitation based on magnetostrictive effect in pipeline site inspections.

Topics: Inspection , Waves
Commentary by Dr. Valentin Fuster
2018;():V01BT01A041. doi:10.1115/PVP2018-84583.

During the final post welding heat treatment (PWHT), residual stress relieves gradually with the accumulation of creep strain. However, reheat cracking with intergranular characteristic will occur when grain boundary cannot accommodate this kind of strain for some special steel welding, such as the welding coarse grained heat affected zone (CGHAZ) of 2.25Cr1Mo0.25V steel. Based on the principle of stress relaxation similar to the process of PWHT, two methods are applied to study the strain criterion of reheat cracking. Stress relaxation testing is performed on CGHAZ materials prepared by Gleeble thermomechanical simulator. The critical strain is calculated using the relationship between stress reduction and creep deformation. Self-loaded notched C-ring specimens are tested taken from the welding structure, coupled with finite element modeling and multiaxial creep coefficient to determine the critical strain. The results show that there is a large numerical difference between the critical strains from two methods. The possible reasons for the difference are given. Regarding the PWHT as a service process, whether the critical strain values obtained exceed the strain limits in ASME-NH is discussed.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A042. doi:10.1115/PVP2018-84694.

High pressure hydrogen storage tank is a key component in hydrogen supply systems of the fuel cell vehicles. Composite hydrogen storage tanks own some advantages such as high stiffness, density and strength, and the composite tanks for storage of 70 MPa hydrogen is the hotspot of research. However, composite hydrogen storage tanks have the potential leakage and explosion hazards in the process of using due to the flammable and explosive storage medium. hydrogen cycle test is an important means to detect the macroscopic strength, safety margin, structure design rationality and reliability of composite hydrogen storage tanks. The relevant standard of 70MPa composite hydrogen storage tanks is not yet introduced in China. At present, the european union, Japan and the international standardization organization have formulated the composite hydrogen storage tanks standard or code, which have an important reference for the formulation of the relevant standards in china. In this study, foreign standards related hydrogen cycling test content such as (EU) 406-2010, ANSI HGV 2-2014, ISO CD 19881-2015, SAE J2579-2013 and ECE/TRANS/180-2013 are systematically studied, which could be helpful for the establishment of relevant standards in China and the development of hydrogen cycle test equipment.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A043. doi:10.1115/PVP2018-84743.

Tube trailers with large capacity gas cylinders have been widely used to deliver compressed natural gas (CNG) in China in recent years. Most of the large capacity cylinders are full steel cylinders (Type CNG-1), and the tube trailers with large capacity glass fiber hoop-wrapped composite cylinders with steel liners (Type CNG-2) have entered into Chinese market since 2011. The tube trailers with CNG-2 cylinders can carry more gases than those with CNG-1 cylinders, but need more loading and cooling time to control their wall-temperatures according to the experiences of on-site operators in gas stations. In this paper, the wall-temperatures of those two types of large capacity cylinders on two tube trailers have been tested using a thermal infrared imaging camera during CNG loading processes in two natural gas stations in China. And numerical simulation on the CNG loading processes has been carried out to investigate the temperature changes of the gases and the shells of the cylinders. The results show that CNG-2 cylinders can have less maximum temperatures of the gases and the shells and less wall-temperature increments at the cost of lower loading rate and longer loading time.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A044. doi:10.1115/PVP2018-84808.

Hydrogen energy as the cleanest fuel to replace gasoline has been accepted by society, hydrogen fuel could be promoted based on the safety of hydrogen-fuel storage containers. For risk-controlling of hydrogen storage containers, there are many laws and regulations in UN and EU set the strict technical requirements on high pressure hydrogen storage systems and require a lot of rigorous experimental verification should be performed before mass production. Frame of GTR No.13, ECER No.134 and EU No406/2010 and the content relevant with high-pressure hydrogen storage container would be discussed emphatically in this paper. Rigorous testing methods in regulations and standards are compared and comments on hydrogen storage container performance testing are provided, besides, some important testing items are discussed.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A045. doi:10.1115/PVP2018-84873.

This paper aims at in-service inspection of a petrochemical companies’ process pipeline which was found to have corrosive pits during a regular periodic inspection. By adopting miniature specimen sampling method, some (size of 12 × 12 × 1.3mm) sheets were obtained in the pipeline. Yield strength and tensile strength of the material can be obtained by small punch test method. Fitness-For-Service (FFS) procedures based on different standards are performed on the corrosive pits and the sampling areas to assess their safety. The research shows that it is feasible to obtain the actual mechanical properties of the material in the field by using the miniature specimen sampling method. The method mentioned in this paper is of guiding significance for FFS assessment on in-service pressure pipelines.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A046. doi:10.1115/PVP2018-84954.

The main failure mode of a large vacuum spherical tank, subjected to uniformly external pressure, is buckling of thin shell. Taking a large vacuum spherical tank with volume of 18000m3 for research objects, the different calculation methods of buckling pressure between the ASME standard, analytic method and finite element method, were compared and analyzed. Using ANSYS software, the buckling pressure of stiffened spherical tanks were analyzed, and the optimum design of the thickness, height and quantity of stiffeners are studied. The results of analysis showed that the anti-instability ability of the large vacuum spherical tank can be improved significantly by a reasonable arrangement of stiffeners.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A047. doi:10.1115/PVP2018-85008.

A failure mode and effect analysis is performed on 58 centrifugal pumps of Atmospheric and Vacuum Distillation Unit in one of the plants of SINOPEC. Firstly, the failure modes of the pumps are classified. Then, the failure causes of the pumps are analyzed. After that, the influence of the failure and the risk level are determined. Finally, the maintenance strategies to reduce risk are put forward according to the equipment failure mode and risk level, which is the basis for the maintenance of pumps. The results show that the main failure modes of the centrifugal pumps are process medium leakage, unexpected stop, vibration, damage, noise, low output and of useful medium leakage. The mean time to failures of centrifugal pumps is 14533 hours, about 20 months. Nine centrifugal pumps are in the risk of high or medium-high, which accounts for 15.52% of the total number of pumps.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in European Codes and Standards

2018;():V01BT01A048. doi:10.1115/PVP2018-84038.

Nuclear power plants contain certain components whose gross failure would lead to intolerable radiological consequences. In the UK, a common terminology used for such components is Very High Integrity (VHI). If it is not possible to engineer lines of protection for these components, a safety case must demonstrate to UK regulators that the probability of gross failure is demonstrably so low that it can be discounted. A previous paper [Ref. 1] has described, at a high level, how the structural integrity safety case for a nuclear new build project in the UK — the UK Advanced Boiling Water Reactor (UK ABWR) is being structured.

As described in [Ref. 1], the structural integrity safety case for the UK ABWR is based on the guidance provided by the UK Technical Advisory Group on Structural Integrity (TAGSI) and aims to demonstrate a multi-legged safety case with robust and independent legs giving confidence of defense in depth. Design to the internationally recognized ASME code [Refs. 2, 3, 4] is supplemented by a significant number of beyond code requirements such as supplementary inspection and inspection qualification, augmented material testing requirements, defect tolerance assessment to the well-established R6 procedure [Ref. 5], and demonstration that design and manufacturing processes have reduced risks to As Low as Reasonably Practicable (ALARP).

This paper provides an updated position of the progress made on the UK ABWR project. It also provides more specific details on the activities the future licensee, Horizon Nuclear Power, has performed in support of the demonstration that design and manufacturing processes have reduced risks to ALARP. This kind of additional work is vital to providing the UK regulator with confidence that the risk of failure of VHI components has been reduced to ALARP.

Topics: Safety , Reliability
Commentary by Dr. Valentin Fuster
2018;():V01BT01A049. doi:10.1115/PVP2018-84119.

Fitness for service assessment procedures rely on flaw interaction rules for assessment of multiple flaws in close proximity. Such rules are aimed at avoiding excessive amplification of the crack driving force that may result in a non-conservative fracture assessment. In BS7910, the 2013 edition [1] introduced a new flaw interaction rule for the co-planar flaws where the proximity of adjacent flaws is judged based on flaw height (i.e. s = 0.5*max(a1,a2) for surface flaws). The rule was introduced for flaws with aspect ratio of a/c < 1 for both flaws, while for other flaw shapes and combinations the earlier rule from the predecessor document PD6493:1993 [2] was retained. This paper summarises the recent work done by the authors and work from literature to examine the applicability of the s = 0.5*max(a1,a2) rule to flaws with aspect ratio a/c ≥ 1 and dissimilar flaw combinations.

It is shown that the current BS7910 rule based on s = 0.5*max(a1,a2) produces a conservative flaw assessment with the use of BS7910 solutions for stress intensity factor and reference stress. An exception are cases of two deep surface flaws where the rule is proposed to change to:

s ≤ max(a1, a2) for two surface flaws with a1/t & a2/t > 0.5

Commentary by Dr. Valentin Fuster
2018;():V01BT01A050. doi:10.1115/PVP2018-84139.

VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989.

VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes.

Last upgrading and principal extending of this VERLIFE Procedure was realized within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013.

As all versions of the VERLIFE procedure were coordinated by the Czech and first version was based on the Czech version of the NTD ASI, there have been simultaneously incorporated into the Czech NTD ASI (Normative Technical Documentation of the Czech Association of Mechanical Engineers) guidelines that are accepted by the Czech State Office for Nuclear Safety for the use in evaluation of Czech NPPs.

This document has several parts: Section IV – “Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs” deals with the evaluation during NPP operation. Main part of the document is divided into four main parts:

- Evaluation of resistance of components and piping against non-ductile failure

- Evaluation of resistance of component and piping against fatigue damage,

- Evaluation of resistance of components and piping against corrosion-mechanical damage

- Evaluation of residual lifetime of components and piping with defects found during in-service inspections

Additionally, several appendices are included for detailed description of individual parts of evaluation, e.g.

- Determination of neutron fluences in reactor pressure vessel and internals

- Determination of degradation of materials during operation

- Requirements for evaluation of pressurized thermal shock regimes

- Evaluation of corrosion-errosion effects in piping

- Environmental fatigue evaluation

- Evaluation of reactor pressure vessel failure probability

Finally, the following appendices dealing with components integrity have been included:

- Lifetime of reactor pressure vessel internals

- Leak-before-break concept for WWER piping

- No-break-zone for WWER piping

The paper will describe structure and main principles of this Section IV.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A051. doi:10.1115/PVP2018-84409.

The paper will cover the general approach followed by nuclear code RCC-M [1] of AFCEN, the French Society for Design, Construction and In-Service Inspection Rules for Nuclear Island Components, in codes and standards setting, from the technical and organizational points of view. After reminding the main modifications introduced in the 2016 & 2017 editions, the main evolutions expected in the 2018 edition of RCC-M code will be explained and commented, as well as the main new topics of activity of RCC-M subcommittee. The presentation highlights how the industrial experience is currently integrated into the RCC-M code. It also develops how the substantial effort carried out by AFCEN for the last three years, to demonstrate the conformity with European and French regulatory requirements, led not only to the development of dedicated guides and modifications in regulatory appendices of the code, but also to improvements in the main parts of the code.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A052. doi:10.1115/PVP2018-84681.

The United Kingdom (UK) Forum for Engineering Structural Integrity (FESI) is the membership organisation for engineering structural integrity (ESI) in the UK. This paper provides an overview of FESI in terms of summarising its recent activities with particular regard to conferences, workshops/seminars, training courses, and the strengthening of links and ties to other organisational bodies associated with elements of ESI. A commentary is included on how some of these activities relate to codes and standards to meet future demands of existing and future plants for extended periods of operation.

Topics: Ferrosilicon
Commentary by Dr. Valentin Fuster
2018;():V01BT01A053. doi:10.1115/PVP2018-84904.

The European Pressure Vessel Standard EN 13445 provides in its part 3 (Design) a simplified method (Clause 17) and a detailed method for fatigue assessment (Clause 18) of unwelded and welded components. Clause 18 “Detailed Assessment of Fatigue Life” is under principal revision within the framework of the European working group “CEN/TC 54/WG 53 – Design methods” in order to reach a significant increase in user-friendliness and a clear guideline for the application.

This paper is focused on the new recommendations for the thermal fatigue evaluation given in the new informative annex NA “Instructions for structural stress oriented finite element analyses using brick and shell elements”. In this annex NA different application methods for the determination of structural stresses are explained in connection with the requirements for finite element models and analyses.

This paper will give a short overview of the proposed approaches of structural stress determination in the new draft annex NA of the revised EN 13445-3 with special recommendations for thermal fatigue evaluation application. This constitutes an extension of the usual and established application of the structural stress approach for welds subjected to mechanical loading conditions. It will present the current state of the approaches based on the results of fatigue analyses according to EN 13445-3 Clause 18 for two different application examples. For validation purposes, these different approaches are compared with the results of Clause 17 of the EN 13445 and other pressure vessel design codes.

Topics: Fatigue , Stress
Commentary by Dr. Valentin Fuster
2018;():V01BT01A054. doi:10.1115/PVP2018-85075.

This paper provides an overview of the ongoing activities in the UK to understand the possible needs and development opportunities for design codes, standards and assessment procedures when looking at Small Modular Reactors (SMRs) and Generation (Gen) IV reactors. The project (at the time of the conference) is progressing towards the completion of the initial gaps analysis phase of the work. This project is also part of a wider programme of work being supported by the Innovate UK to consider other pertinent aspects such as materials, automated manufacturing, large-scale manufacture and assembly and modularised build. This paper summarises these activities and the findings to-date.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A055. doi:10.1115/PVP2018-85155.

The Standard Technical Documentation of the Association of Mechanical Engineers is elaborated by a team of experts on the base of an actual knowledge and practice as a part of a row of recommendations for an assessment of strength and reliability, choice of materials and solution of service problems of the Czech nuclear power plants. It was elaborated by the Association of Mechanical Engineers as an important help for a practice. This issue is published from the following main reasons:

1. The criteria for an assessment of reliability and safety of technical tasks have changed in a last decades by influence of development of fracture mechanics and a diagnostics of an actual components of the nuclear power plants, exposed to a static and dynamic strength in a service and to effects of an aggressive environment, leading to an origin of integrity failure. The access is changed now from an assessment of an original strength to a long-term operation and service reliability. A conception of a choice of suitable materials has changed, especially about an influence of production technology on barriers against a degradation of material mechanical characteristics. Imaginations also changed about increasing of a resistance against a damage of material components in complex service conditions. New scientific knowledge it is necessary to apply responsibly just in a construction of as important equipment as nuclear power plants are.

2. A social responsibility is increasing for economic and ecological behavior and hence manners of service control are always innovated. A problematic of a residual lifetime of those tasks is connected with. A prolongation of a lifetime without a risk of equipment failure hereto is a principal demand of capital-intensive units.

3. In a last decades a possibility of efficient use either an automatic service diagnostics or an inspection of component state in service brakes has deepened. It is necessary to enable a right engagement of this instrumentation and to interpret in a right way obtained information.

4. An international cooperation in this important region is developed, an elaboration of information and their common deepening has a substantial importance for a future of a nuclear energetic. A dissembling of service obstacles, an objective analysis of failure situations and a method of their control is today’s obvious duty of an international cooperation. Programs of all variants of a possible damage and its control must be prepared.

It is impossible at once to apply all those aspects in provisions and recommendations. Issued provisions however must be constantly improved and complemented.

Commentary by Dr. Valentin Fuster

Codes and Standards: Recent Developments in Japanese Codes and Standards

2018;():V01BT01A056. doi:10.1115/PVP2018-84956.

When cracked bodies are subjected to cyclic loading, fatigue crack growth evaluation is often required from the viewpoint of the assurance of fitness-for-service. For cyclic loading with constant amplitude, crack growth can be calculated by integrating a fatigue crack growth rate law provided by the Paris law in terms of stress intensity factor range. However, the cause of cyclic loading supposed in actual structures is not simple. For an example of LWR (Light Water Reactor) plants, a number of datasets of cyclic loading with different amplitudes is necessary for specific transient events. And the chronological order of individual transients cannot be determined. In this paper, a universal procedure to deal with multiple transients was developed in case that the chronological order of the transients was indefinite. Cyclic loading sequence such that the loadings were lined up in the order corresponding to larger amplitude gave the most conservative crack growth prediction among the possible sequences from the set of the specific transient events. Nevertheless the effect of the sequence was quite limited and the differences in fatigue crack growth were much less than the accuracy of the analysis outputs.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A057. doi:10.1115/PVP2018-84965.

In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency for structural integrity assessment of reactor pressure vessels (RPVs). The most recent release is PASCAL Version 4 (hereafter, PSACAL4) which can be used to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and pressurized thermal shock events. For the integrity assessment of RPVs, development of crack evaluation models is important. In this study, finite element analyses are performed firstly to verify the stress intensity factor calculations of cracks in PASCAL4. In addition, the applicability of the crack evaluation models in PASCAL4 such as the location of embedded cracks, crack shape and depth of surface cracks, and the increment of crack propagation is investigated. Based on sensitivity analyses of crack evaluation models for Japanese RPVs using PASCAL4, the effects of these evaluation models on failure frequency are clarified. From the analysis results, crack evaluation models recommended to the failure frequency evaluation for a Japanese model RPV are discussed.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A058. doi:10.1115/PVP2018-84989.

Reactor pressure vessels (RPVs) in nuclear power plants are important components that prevent non-ductile fracture considering neutron irradiation embrittlement as aging degradation and several types of transients. An analysis code called PASCAL for assessing failure frequencies of RPVs based on probabilistic fracture mechanics has been developed by the Japan Atomic Energy Agency. In failure frequency analyses, flaw size distribution in RPVs is one of the most important parameters, and it is determined by considering possible flaws generated during fabrication and the flaw-detection capabilities of nondestructive examinations (NDEs). Flaw-detection capabilities of NDEs are represented as probability of detection (POD) curves related to flaw sizes.

In this study, the effects of NDEs on failure frequencies of RPVs are evaluated using PASCAL considering simplified POD curves in terms of minimum detectable flaw size, the smallest probability of non-detection (PND), and flaw size where POD value reaches the smallest PND.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A059. doi:10.1115/PVP2018-84995.

This paper describes the requirements for securing the reliability of pipes with a circumferential surface flaw when they are used continually without repair. In the Rules on Fitness-for-Service (the Rules on FFS) of the Japan Society of Mechanical Engineers (JSME), limitations for flaw angle and flaw depth, and the safety margin for fracture evaluation are prescribed to ensure structural integrity during the evaluation period. In addition, for continued use without repair, three successive inspections are conducted and the prediction of flaw size is reviewed. The influence of these rules on the fracture probability when applying the limit load analysis was evaluated. From these evaluations, the flaw depth limitation and improvement of the safety factor are ineffective in improving the reliability of fracture evaluation, and the most influential factor is the prediction accuracy of flaw size. To improve the prediction accuracy of flaw size, it is important to review the predicted flaw size to envelop its actual size at the end of the evaluation period using the results of successive inspections. Based on this examination, the framework of the rules of the successive inspections and flaw evaluation incorporating the revision of the flaw size at the end of the evaluation period was proposed.

Commentary by Dr. Valentin Fuster

Codes and Standards: Repair, Replacement, and Mitigation for Fitness-for-Service Rules

2018;():V01BT01A060. doi:10.1115/PVP2018-84472.

Cracking on metal surface including Stress Corrosion Cracking (SCC) is a concern for nuclear power plant reactor system piping and components. Several repair techniques have been developed over the years that help address the cracking issues. In the United States, some crack remaining repair technologies such as Weld Overlay (WOL) technique have already been developed and also some technologies such as Excavate and Weld Repair (EWR) are still being actively discussed. Fitness-for-Service (FFS) codes in Japan also define some repair methods which allow repairing the components without removing the cracks. This paper introduces three repair methods which are described in the Fitness-for Service code in Japan, the WOL, the cap repair technique for Bottom Mounted Instrumentation (BMI) nozzles and seal welding.

WOL is a repair method that maintains structural integrity by depositing Intergranular Stress Corrosion Cracking (IGSCC) resistant weld to the outer surface of cracked circumferential welded joints in primary piping made of austenitic stainless steel. The cap repair technique for BMI nozzles is a method of repairing from the outer surface when leakage due to SCC occurs. The cap develops a new boundary outside the BMI nozzle to prevent the leakage by retaining the leakage inside the cap. The seal welding technique is also a repair method to cover the cracks by welding to prevent propagation of the cracks and to prevent leakage of reactor water.

Provided the cracks are adequately managed and evaluated, these crack remaining repair methods are very beneficial.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A061. doi:10.1115/PVP2018-84480.

The process from the engineering development to actual application of the new techniques used for the maintenance activities for Japanese nuclear power plants’ (NPPs) component in operation, the JANTI Guideline, “Guideline on Application Process of Techniques Developed for Maintenance Activities” was developed to propose clarification activities to be checked or examined in each stage throughout process. The guideline focuses on techniques used for especially such remedial activities as repair, replacement and mitigation of ageing. In engineering development stage, all information on the new techniques are provided to be checked or examined at each stage of the process followed further, such as code/standard formulation stage. It is important to carry out share recognition for the activities, which should be performed in each process in the prompt application to the actual plant system efficiently, among the persons or organizations concerned.

Topics: Maintenance
Commentary by Dr. Valentin Fuster
2018;():V01BT01A062. doi:10.1115/PVP2018-84512.

Maintenance of nuclear power plant facilities involves activities comprising a large system composed of both plant hardware and human subsystems to assure safe and reliable operation. Maintenance activities are composed of inspection, evaluation and corrective measures. Corrective measures are countermeasures for aging degradation, e.g., resetting the inspection period based on the results of inspection and evaluation; mitigation of degradation phenomenon; repair or replacement; preventive maintenance; etc. The corrective measures merit special attention as they are important and valuable actions in order to promote continued efficient and safe plant operations. It is necessary to develop a set of regulatory and industrial technical requirements for a well-structured, documented set of standards, so that corrective measures can be used and applied uniformly and effectively. Currently the code and standard system is less developed in Japan than in the United States.

In this study, the authors considered the relationship between degradation and maintenance and the difference of performance requirements between the plant construction stage and the in-service stage. This effort is intended to clarify the issues of regulation for maintenance activities, with an objective to help develop structured regulatory/industrial requirements with a code and standards consistent with appropriate corrective measures.

The Nuclear Regulation Authority (NRA), the regulatory body in Japan, has reviewed the present Japanese inspection system in response to suggestions from the Integrated Regulatory Review Service (IRRS) mission established by International Atomic Energy Agency (IAEA). The NRA has also been developing a new regulatory inspection system similar to the Reactor Oversight Process (ROP) used in the United States. The expectation for the new Japanese inspection system is to focus regulations on plant issues with higher risk importance, considering both plant hardware and human subsystems. The new Japanese regulatory system addressing maintenance is also expected to enhance electric utilities ability to assure safety is self-motivated and sustained.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A063. doi:10.1115/PVP2018-84779.

Minor leakage in brazed joints is occasionally detected in nuclear power plant Class 3 piping systems such as emergency service water cooling lines for equipment room and containment area coolers. Detection of such leakage often causes the system to be taken out of service for performance of repair/replacement activities to restore the piping to its original condition. This can lead to plant shutdowns or outage delays which are very costly, and often at increased safety risk, to accomplish the repairs. To avoid such costs and increased risk, a method is needed to establish structural adequacy of leaking brazed joints so that leakage can be controlled and monitored until the next planned opportunity for replacement of the brazed joint.

During industry testing performed in the late 1950s, over 1200 lap joint tensile test specimens were brazed in a round-robin series of tests performed by 10 laboratories. The results of these tests showed that very little overlap is needed in a brazed lap joint to obtain full strength in the joint. In fact, braze metal shear strength exceeded piping collapse strength in all cases when the braze overlap was only 2.3 times the thickness of the members being brazed. Similar testing was performed by the U.S. Navy, and resulted in development of a NAVSEA document that established a conservative percentage of bond required for MIL F-1183 brazed fittings to prevent structural failure of a leaking brazed joint.

This paper discusses an ASME Section XI Code Case that has been developed to incorporate the results of this testing into a methodology for temporary acceptance of leakage of brazed lap joints in copper, copper-nickel, and nickel-copper ASME Code Class 3 nuclear piping systems.

Topics: Leakage
Commentary by Dr. Valentin Fuster
2018;():V01BT01A064. doi:10.1115/PVP2018-84897.

Cathodic protection (CP) is one of the primary methods to protect buried piping and pressure components from corrosion and is a critical element in asset management of buried piping at nuclear power plants. Implementation of cathodic protection requires non-structural attachments to the buried piping for electrical leads and connections. The method of attaching copper-copper alloy CP leads to carbon steel piping and components using traditional arc welding processes can be difficult and time consuming. A two-step process is frequently used where a carbon steel weld tab is first welded to the pipe or component by a traditional arc welding process. The copper-copper alloy CP lead is then joined to the carbon steel weld tab by the exothermic welding process.

An alternative to this cumbersome two-step process is pin brazing which is an automatic brazing process that uses electric current resistance to heat the interface between a pin capsule and the component. An arc between the pin capsule and the outside surface of the electrical connector is then used to melt the capsule or pin that contains the brazing filler metal. The process is similar to stud welding in that the brazing pin is loaded into an automatic pistol and the brazement is made when the trigger is pulled. ASME Section XI Code Case N-882 delineates rules and requirements for application of pin brazing on Class 2 and 3 pressure boundary components. This paper provides the background and description of the pin brazing process with a summary of the technical basis for Case N-882.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A065. doi:10.1115/PVP2018-85154.

This report introduces techniques involving novel calculation methods developed by the Electric Power Research Institute and Bechtel Marine Propulsion Corporation [1] that can be used to develop an improved understanding of welding factors affecting substrate weld heat input and dilution. These improved calculation methods are especially warranted in ASME BPVC Section XI repair and replacement welding activities such as mitigation of hot cracking (dilution control) and optimization of temper bead welding (heat input control). Various parameter sets are tested to demonstrate the effectiveness of the new calculation methods and how they can be refined in future studies to further improve accuracy. Additionally, this report highlights the limitations associated with the ASME BPVC Section IX theoretical heat input and power ratio calculation methods widely used in the welding industry for control of substrate heat input and weld dilution. Weld trials are performed in this study to demonstrate the limitations and determine proper precautions to take when using the calculation methods.

Topics: Heat
Commentary by Dr. Valentin Fuster

Codes and Standards: Structural Integrity of Pressure Components

2018;():V01BT01A066. doi:10.1115/PVP2018-84195.

The high-pressure applications of traditional Gasket Plate Heat Exchangers (GPHE) are limited by the sheet material, gasket material and the design of the GPHE. The newly developed stainless steel grade UNS S82031 (EN 1.4637) is a duplex stainless steel grade with improved formability compared to other duplex stainless steel grades. It can be used in forming intensive applications, such as in typical chevron corrugated-plate heat exchangers. Compared to the standard austenitic stainless steel grades typically used, the new duplex stainless steel increases the application performance of GPHE applications, such as maintaining higher working pressures.

In this paper, the properties and material behavior of UNS S82031 is compared with the standard austenitic stainless steel UNS S30400 (EN 1.4301) in a pressure test at different working pressures. The pressure test is used to evaluate the performance of the selected material for a GPHE application. The experimental test results are compared with Finite Element (FE) simulation of the same pressure test, to achieve a deeper understanding of the results. The results show that plates made of UNS S82031 can withstand significantly higher working pressures than UNS S30400 for the same PHE-design, proving the new duplex stainless steel UNS S82031 is more suitable for high pressure GPHE applications.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A067. doi:10.1115/PVP2018-84279.

A significant number of stainless steel components within the boilers of the UK advanced gas-cooled reactor (AGR) plants are subjected to oxidation, carburisation and other changes in the underlying microstructure of the material during operation. This results from exposure to the pressurised CO2-based primary circuit coolant at temperatures from about 500 to 650°C. It is believed that there is a synergistic relationship between the pressurised CO2 coolant environment and creep-fatigue initiation and cracking. Devising and implementing an evaluation methodology to account for oxidation and carburisation to enable conservative lifetime assessments is essential to current and future plant safety. Therefore, the development of a new and fundamental understanding of environmentally assisted degradation and failure mechanisms is required. It has been postulated that the mechanism underlying the initiation of cracks is carburisation associated with the presence of a duplex oxide layer. In this study, the material-environment interaction for Type 316H stainless steel under simulated AGR conditions has been investigated to increase the understanding of the combined effects of stress, strain and surface preparation, for example, on oxidation and cracking behavior. Experimental data are presented which show that surface deformation promotes the formation of a thin, protective oxide scale, which does not protrude along the grain boundaries, whereas a deformation-free surface leads to the formation of thick duplex oxide layers as well as intergranular oxide penetration. Furthermore, an increased surface hardness due to carburisation has been observed for the undeformed surface only, suggesting that carburisation occurs at an early stage on a chemically treated surface. It is found that when the substrate is plastically deformed and under the effect of active stress, the thin oxide on the mechanically deformed surface can be disrupted, resulting in similar behaviour to a chemically treated surface with no deformation.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A068. doi:10.1115/PVP2018-84345.

Since the implementation of pressure-temperature (P-T) limit curves in the 1960s for light water reactors, the P-T limit curves have been based on the limiting locations in the reactor coolant system, which are typically the irradiated reactor pressure vessel (RPV) region adjacent to the core (beltline) and the closure head flange. Recently, it has been questioned as to whether the reactor vessel inlet or outlet nozzle corners could be more limiting due to the stress concentration at these locations. The discussion presented in this paper provides technical justification that the RPV nozzle corner P-T limit curves are bounded by the traditional P-T limit curves for the pressurized water reactors (PWRs).

The current approach in evaluating the Pressurized Water Reactor Inlet and Outlet nozzle corner regions with respect to plant heatup and cooldown Pressure Temperature Limit Curves contains a number of conservatisms. These conservatisms include postulation of a large 1/4T flaw at the nozzle corner region, use of RTNDT (reference nil-ductility temperature), and fracture toughness prediction based on plane strain fracture toughness. The paper herein discusses several factors that can be considered to improve the pressure temperature limit curves for nozzle corners and increase the operating window for nuclear power plant operations.

Prior to the 2013 edition, the ASME Section XI Appendix G did not require the use of a 1/4T flaw for the nozzle corners; furthermore, a smaller postulated flaw size is permissible. Based on inspection capability and experience, a smaller flaw size can easily be justified. The use of a smaller flaw size reduces the stress intensity factors and allows for the benefit of being able to take advantage of increased material toughness due to the loss of constraint at the nozzle corner geometry. The analysis herein considers the calculation of stress intensity factors for small postulated nozzle corner flaws based on a 3D finite element analysis for Westinghouse PWR inlet and outlet nozzle corner regions. The Finite Element Analysis (FEA) stress intensity factors along the crack front are used in the determination of allowable pressures for the cooldown transient Pressure-Temperature limit curves.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A069. doi:10.1115/PVP2018-84412.

This paper presents an overview of added features in a new version of the FAVOR (Fracture Analysis of Vessels Oak Ridge) computer code called FAVOR-OCI. The original FAVOR code was developed at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission (NRC). FAVOR is applied by US and international nuclear power industries to perform deterministic and probabilistic fracture mechanics analyses of commercial nuclear reactor pressure vessels (RPVs). Applications of FAVOR are focused on insuring that the structural integrity of aging, and increasingly embrittled, RPVs is maintained throughout their licensed service life. Based on the final ORNL release of FAVOR, v16.1, FAVOR-OCI extends existing deterministic features of FAVOR while preserving all previously-existing probabilistic capabilities of FAVOR.

The objective of this paper is to describe new deterministic features in FAVOR-OCI that can be applied to analytical evaluations of planar flaws. These evaluations are consistent with the acceptance criteria of ASME Code, Section XI, Article IWB-3610, including Subarticles IWB-3611 (acceptance based on flaw size) and IWB-3612 (acceptance based on applied stress intensity factor). The linear elastic fracture mechanics (LEFM) capabilities of FAVOR-OCI also incorporate the analytical procedures presented in the Nonmandatory Appendix A, Analysis of Flaws, Article A-3000, Method of KI Determination, for both surface and subsurface (embedded) flaws.

The paper describes a computational methodology for determining critical values of fracture-related parameters that satisfy ASME Code Section XI acceptance criteria for flaws exposed to multiple thermal-hydraulic transients. These compute-intensive analyses can be carried out with a single execution of FAVOR-OCI. The new methodology determines critical values by solving for either the point of tangency or point of intersection between applied KI versus time histories and a user-selected cleavage initiation toughness material property (e.g., ASME KIc, FAVOR Weibull KIc, or Master Curve Weibull KJc) for surface or subsurface flaws. Situations where warm prestress conditions apply can also be addressed. The paper highlights a need for this new capability via applications to a recent independent review of safety cases for RPVs in two Belgian nuclear power plants (NPPs). That review required ASME Section XI assessments of several thousand embedded, quasi-laminar flaws in the wall of each RPV Analysis results provided by the new capability contributed to the technical bases compiled from several sources by the Belgian nuclear regulatory agency (FANC) and eventually used by FANC to justify the restart of these NPPs.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A070. doi:10.1115/PVP2018-84582.

Clause 4.5 of ASME Section VIII Division 2[1] provides rules for compensation of openings in cylindrical shells having fitted nozzles.

The rules provided in Clause 4.5.5 of ASME Section VIII Division 2[1] are based on pressure-area method which is based on ensuring that the reactive force provided by the vessel material is greater than or equal to the load from the pressure.

Clause 3.5.4 of PD 5500[5] provides rules for compensation of opening and nozzle connections.

Clause 3.5.4.3 provides requirements for the design of isolated openings and nozzle connections in the form of design procedure. Clause 3.5.4.4 provides requirements for groups of openings and the procedure allows the checking of chosen geometry.

Clause 3.5.4.9 of PD 5500[5] provides rules for compensation of openings by pressure-area method to those geometries which confirms to the geometric limitations specified therein. This method has extensive satisfactory use in European Code of practice and has been adopted in BS EN 13445-3 also.

The key element in applying the pressure area method is to determine the dimensions of the reinforcing zone, i.e., the length of the shell, height of the nozzle and reinforcing pad dimensions (if reinforcing pad is provided), that resist the applied pressure.

In comparison to certain restrictions in PD 5500[5] there appears to be no restriction on the physical dimensions of the nozzle or shell in ASME Section VIII Division 2[1], as long as the required area AT is obtained and the stresses are within allowable limits.

It is therefore possible that all of the required area AT is obtained either from the nozzle or from the shell. While both these alternatives would be acceptable in ASME Section VIII Division 2[1] design, the actual stresses at the shell/nozzle junction may vary considerably.

The work reported in this paper — a comparative study of pressure area method of nozzle compensation in ASME Section VIII Division 2[1] and PD 5500[5] for restrictions in nozzle dimensions was undertaken to compare the results obtained from both the Codes and is an extension of work carried out and published as PVP2015-45564.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A071. doi:10.1115/PVP2018-84639.

To investigate the fatigue behavior of metastable austenite steels in the VHCF-regime, high loading frequencies are essential to realize acceptable testing times. Hence, two high-frequency testing systems were used at the authors’ institute: an ultrasonic testing system with a test frequency of 20 000 Hz and also, a servohydraulic system with a test frequency of 980 Hz. In the present study, two different batches of the metastable austenitic stainless steel AISI 347 were investigated.

Fatigue tests on metastable austenitic steel AISI 347 batch A were carried out at an ultrasonic test system at a test frequency of 20 000 Hz, at ambient temperature. Because the test rig acts as a mechanical resonant circuit excited by a piezoelectric transducer the specimen must be designed for oscillation in its vibration Eigenmode at the test frequency to assure maximum displacement at the end and maximum stress in the gauge length center, respectively. For analyzing the deformation behavior during the tests, the change in temperature was measured. Additionally, Feritscope™ measurements at the specimen surface were performed ex-situ after defined load cycles. First results showed a pronounced development of phase transformation from paramagnetic face-centered cubic γ-austenite to ferromagnetic body-centered cubic α‘-martensite. Because formation of α‘-martensite influences the transient behavior and high frequency loadings leads to pronounced self-heating of the material, ultrasonic fatigue tests on metastable austenites represent a challenge in controlling of displacement amplitude and limiting the specimen temperature.

First investigations on metastable austenitc steel AISI 347 batch B using a servohydraulic test system at a frequency of 980 Hz and a temperature of T = 300 °C resulted in no fatigue failure beyond N = 107 cycles in the VHCF-regime. However, only specimens with a low content of cyclic deformation-induced α‘-martensite achieved the ultimate number of cycles (Nu = 5·108).

Commentary by Dr. Valentin Fuster
2018;():V01BT01A072. doi:10.1115/PVP2018-84662.

Joint design, preheating, post-weld-heat treatment and operator ability are key factors for the occurrence of critical residual stresses in the welds. In particular, they may affect creep life results of the analysis: to evaluate this, their characterization requires modeling welding process, such as that carried out in the present work for ASTM A 335-Grade P22 weld. It is a weld of the two analyzed in previous work on the high-temperature-section (superheater/reheater) lower headers of the bottom-supported heat-recovery steam generator (HRSG). Present study includes modeling the only weld-lay, gas tungsten arc, manual, for the finned-tube angle joint to the cylinder. It is the same as the first of the three weld-lays for the other previously analyzed, end-plug circumferential “V”-groove butt joint. The material considered for base metal is 2¼Cr 1Mo forged, normalized, tempered with creep strain rates higher than the weld’s as it appeared from analysis of previous work. Presently, the study first compares max tangential stress evaluated by thermal analysis on the circumferential weld with the average applied normally on the joint, sustained case. Roughly, the latter one has nearly same importance as axial membrane stress on the cylinder wall for the end-plug butt joint, pressure case; it is smaller for the finned-tube angle joint. Then, study compares previous creep results obtained considering SRF = 0.9 for the finned-tube joint weld with those counting the residual-stress increase in stress analysis (different creep law’s coefficients). Finally, study compares creep results for the finned-tube joint weld obtained with and without residual-stresses (same creep power-law). The objective is to comprehend residual-stress influence over creep-redistribution and creep relaxation.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A073. doi:10.1115/PVP2018-85106.

A recent surge of nuclear industry interest in the development of smaller, modular/factory-built nuclear reactors has been underway for the last few years. Recently, the Canadian Nuclear Safety Commission (CNSC), which is the Canadian nuclear regulatory authority, has been reviewing a number of proposed Small Modular Reactor (SMR) designs through the pre-licensing Vendor Design Review (VDR) process. Some of these SMR designs propose to operate under relatively high neutron fluxes and at elevated temperatures to achieve high thermal efficiency compared to current operating power reactors. Other designs employ a highly corrosive coolant medium. Such operating conditions could accelerate or present unforeseen degradation mechanisms that could adversely affect the integrity and service life of the pressure boundary. It is therefore crucial to understand the material behavior under these conditions to devise appropriate design rules for metallic and non-metallic components. Thus, a systematic approach and a well-designed research and development program to address any knowledge gaps in material behavior at the early design stage for a SMR is crucial, so that potential/plausible degradation mechanisms are accounted for and the structural integrity of Pressure Retaining Systems and Components (PRSC) is maintained over the design life. This paper discusses a proposed framework based on key review findings from the pre-licensing VDR of several SMR designs in Canada.

Commentary by Dr. Valentin Fuster

Codes and Standards: Technical Harmonization and Emerging Codes and Standards

2018;():V01BT01A074. doi:10.1115/PVP2018-84549.

In ASME VIII-2 Code 2004 and previous versions, evaluation procedures for thermal stress ratcheting were provided, which is substantially based on Bree’s diagram. According to VIII-2-2004, the application scope of this method is limited to “in shell” and “general membrane stress due to pressure”. In ASME VIII-2 Code 2007 and later versions, the limitation of “in shell” has been removed and “general membrane stress” was extended to “general or local primary membrane equivalent stress”. However, these methods have their own limitations, and consequently can’t apply to the engineering project widely and rapidly. Rapid thermal stress ratcheting assessment calculations for pressured components with arbitrary geometries and discontinuity effects based on the Linear Matching Method (LMM) were available. Several cases are provided to show how to establish shakedown and ratchet boundaries quickly and easily for components subjected to arbitrary combinations of thermal-mechanical loads. The Bree-like ratchet boundaries for several typical components are obtained, which is useful and valuable for engineering design in Design-by-Analysis (DBA) for pressured components.

Commentary by Dr. Valentin Fuster

Codes and Standards: Uncertainty Characterization in Probabilistic Assessments of Structural Integrity

2018;():V01BT01A075. doi:10.1115/PVP2018-84963.

A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency based on Japanese methods and data to evaluate failure probabilities and failure frequencies of Japanese reactor pressure vessels (RPVs) considering pressurized thermal shock (PTS) events and neutron irradiation embrittlement. To verify PASCAL, we have been performing benchmark analyses by using a PFM code FAVOR which has been developed in the United States and utilized in nuclear regulation. Based on two-year activities, the applicability of PASCAL in failure probability and failure frequency evaluation of Japanese RPVs was confirmed with great confidence. The analysis conditions, approaches and results are given in this paper.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A076. doi:10.1115/PVP2018-85010.

Canadian Nuclear Standard CSA N285.8, “Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU® reactors”(1), permits the use of probabilistic methods when performing assessments of the reactor core. A non-mandatory annex has been proposed for inclusion in the CSA Standard N285.8, to provide guidelines for performing uncertainty analysis in probabilistic fitness-for-service evaluations within the scope of this Standard, such as the probabilistic evaluation of leak-before-break. The proposed annex outlines the general approach to uncertainty analysis as being comprised of the following major activities: identification of influential variables, characterization of uncertainties in influential variables, and subsequent propagation of these uncertainties through the evaluation framework or code.

The application of the proposed guidelines for uncertainty analysis was exercised by performing a pilot study for one of the evaluations within the scope of the CSA Standard N285.8, the probabilistic evaluation of leak-before-break based on a postulated through-wall crack. The pilot study was performed for a representative CANDU reactor unit using the recently developed computer code P-LBB that complies with requirements of Canadian Nuclear Standard N286.7 for quality assurance of analytical, scientific, and design computer programs for nuclear power plants. This paper discusses the approach used and the results obtained in the first stage of this pilot study, the identification of influential variables.

The proposed annex considers three approaches for identifying influential variables, which may be used separately or in combination: analysis of probabilistic evaluation outputs, sensitivity analysis and expert judgment. In this pilot study, local sensitivity analysis was used to identify and rank the influential variables. For each input variable in the probabilistic evaluation of leak-before-break, the local sensitivity coefficient was determined as the relative change in the output variable associated with a relative change of a small magnitude in the input variable. Each input variable was also varied across a large range to assess the linearity of the relationship between the input variable and the output variable. All relevant input variables were ranked according to the absolute value of their sensitivity coefficients to identify the influential variables. On the basis of the results obtained, the pressure tube wall thickness was found to be the most influential variable in the probabilistic evaluation of leak-before-break based on a postulated through-wall crack, followed by the fracture toughness of Zr-2.5Nb pressure tube material and the pressure tube inner diameter. The results obtained at this stage were then used at the second stage of this pilot study, the uncertainty characterization of influential variables, as discussed in the companion paper PVP2018-85011.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A077. doi:10.1115/PVP2018-85011.

Canadian Nuclear Standard CSA N285.8, “Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU® reactors”(1), permits the use of probabilistic methods when assessments of the reactor core are performed. A non-mandatory annex has been proposed for inclusion in the CSA Standard N285.8 to provide guidelines for performing uncertainty analysis in probabilistic fitness-for-service evaluations within the scope of this Standard, such as the probabilistic evaluation of leak-before-break. The proposed annex outlines the general approach to uncertainty analysis as being comprised of the following major activities: identification of influential variables, characterization of uncertainties in influential variables, and subsequent propagation of these uncertainties through the evaluation framework or code. The proposed methodology distinguishes between two types of non-deterministic variables by the method used to obtain their best estimate. Uncertainties are classified by their source, and different uncertainty components are considered when the best estimates for the variables of interest are obtained using calibrated parametric models or analyses and when these estimates are obtained using statistical models or analyses.

The application of the proposed guidelines for uncertainty analysis was exercised by performing a pilot study for one of the evaluations within the scope of the CSA Standard N285.8, the probabilistic evaluation of leak-before-break based on a postulated through-wall crack. The pilot study was performed for a representative CANDU reactor unit using the recently developed software code P-LBB that complies with the requirements of Canadian Nuclear Standard CSA N286.7 for quality assurance of analytical, scientific, and design computer programs for nuclear power plants. This paper discusses the approaches used and the results obtained in the second stage of this pilot study, the uncertainty characterization of influential variables identified as discussed in the companion paper presented at the PVP 2018 Conference (PVP2018-85010).

In the proposed methodology, statistical assessment and expert judgment are recognized as two complementary approaches to uncertainty characterization. In this pilot study, the uncertainty characterization was limited to cases where statistical assessment could be used as the primary approach. Parametric uncertainty and uncertainty due to numerical solutions were considered as the uncertainty components for variables represented by parametric models. Residual uncertainty and uncertainty due to imbalances in the model-basis data set were considered as the uncertainty components for variables represented by statistical models. In general, the uncertainty due to numerical solutions was found to be substantially smaller than the parametric uncertainty for variables represented by parametric models, and the uncertainty due to imbalances in the model basis data set was found to be substantially smaller than the residual uncertainty for variables represented by statistical models.

Commentary by Dr. Valentin Fuster

Codes and Standards: Use of Modern FEA Methods for Code Assessment

2018;():V01BT01A078. doi:10.1115/PVP2018-84035.

The reciprocal effect between a quarter-circle corner crack and a non-aligned surface crack of comparable size is addressed in the present study. The significance of understanding the reciprocal effect between the non-aligned parallel cracks is to assist in the evaluation of non-aligned multiple cracks as required in various fitness-for-service codes. For non-aligned parallel cracks, on-site inspection needs to decide whether the cracks should be treated as coalesced or separate multiple cracks. In the existing literature, criteria and standards for the adjustment of multiple non-aligned cracks are very source dependent, and those criteria and standards are often derived from on-site service experience without rigorous and systematic verification. Based on this observation, the authors previously reported on the effect of an embedded parallel crack on an edge crack in 2-D scenarios and, more recently, in 3-D scenarios of a circular corner crack influenced by a parallel surface crack. It may be just as important to evaluate the mutual effect of a quarter-circle corner crack on a non-aligned surface crack as reflected in their stress intensity factors (SIFs). In the present study, the quarter-circle corner crack and the non-aligned surface crack are assumed to be of the same length a2 = a1 = 15mm. While keeping throughout the entire analysis the geometry of the quarter-circle corner crack unchanged, the relative depth of the semi-elliptical surface crack is varied so that b1/a1 = 0.2–1.0. For each particular case a pair of horizontal (H) and vertical (S) separation distances between the two cracks is chosen (H/a2 = 0.4–2 and S/a2 = −0.5–2) and the SIFs along the 3D crack fronts are extracted for both the corner and the surface crack. The reciprocal effect on the SIFs for both cracks are discussed. It is found that the mutual influence between the corner crack and the surface crack are equally important, and each may dominate the decision making based on present criteria and standards in Fitness-for-Service.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A079. doi:10.1115/PVP2018-84055.

The Pressurized Thermal Shock (PTS) is one of the transients of concern in a nuclear reactor. The phenomenon takes place when the down-comer and internal Reactor Pressure Vessel (RPV) surfaces are rapidly cooled down while the system is still under pressure.

This event is considered life limiting for the safety case where cracks are postulated and the RPV material embrittled by neutron radiation damage.

During this study, an elliptical embedded crack was postulated in a cylindrical geometry simulating the RPV. The system, initially at high temperature and under pressure load condition, was subject to a PTS transient. Thermal and structural analyses were carried out for various crack configurations. Particularly, besides the axial crack, skewed embedded axial cracks were modeled and analyzed (30°, 45° and 65°) for PTS phenomenon. Finite element analysis were carried out with ANSYS considering two variants: no-heat transmission and total heat transmission through the crack. Results showed that the highest stress intensity factor is obtained at an axial crack (0°). Moreover, it is shown that considering no-heat transmission through the crack is a conservative approach.

The ASME code approach to the skewed axial cracks characterization by projection is considered. It is found that the ASME approach is conservative, with an excess conservatism for largely skewed axial cracks. The same conclusions were reached when considering the projection rules from the R6 code.

Commentary by Dr. Valentin Fuster
2018;():V01BT01A080. doi:10.1115/PVP2018-84934.

Class 1 nuclear vessel heads with ellipsoidal or torispherical geometries offer several advantages over spherical heads. They shorten the overall height of the vessel, reduce internal volume, lower the vessel center of gravity, and are more economical. At present Section III only offers tentative thickness formulas for spherical heads in NB-3324 [1]. This paper proposes the addition of tentative thickness formulas for Class 1 torispherical and ellipsoidal heads. A large family of finite element analyses over a practical range of vessel geometries and pressures is used to establish the new thickness equations.

Commentary by Dr. Valentin Fuster

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In