The ASME Section XI Working Groups on Operating Plant Criteria (WGOPC) and on Flaw Evaluation (WGFE) have undertaken an effort to develop a new Code Case (Record # 19-1113) to address the following inquiry: What method(s) are acceptable for obtaining RT NDT , RT To and/or T 0 that account for embrittlement for analytical evaluations performed in accordance with Nonmandatory Appendices A, E, G, or K in lieu of the current requirements of these Appendices? Accounting for embrittlement in Code calculations in an appropriate manner is a critical element to the operating safety of nuclear power plants. The intent of responding to this inquiry is to ensure that Code guidance on this matter is comprehensive, up-to-date with the current state of knowledge and applications, and appropriate for international use. The work on the Code Case aims to unify guidance on the following topics throughout the Code, and to fill gaps as needed: • Source of embrittlement data, • Forecasting of embrittlement trends, • Accounting for embrittlement in the inter-relationships between various toughness properties, and • Accounting for uncertainties associated with embrittlement. This paper outlines the overall structure of the Code Case, and summarizes the progress achieved to date.

The INCEFA+ project (INcreasing Safety in nuclear power plants by Covering gaps in Environmental Fatigue Assessment) is a five year project supported by the European Commission HORIZON2020 programme, which will conclude in June 2020. This project aims to generate and analyse Environmental Assisted Fatigue (EAF) experimental data (approximately 230 fatigue data points generated on austenitic stainless steel), and focuses on the effect of several key parameters such as mean strain, hold times and surface finish, and how they interact with environmental effects (air or PWR environment). This work focuses on the analysis of the data obtained during the INCEFA+ project. More specifically, this paper discusses how the outcome of this analysis can be used to evaluate existing fatigue assessment procedures that incorporate environmental effects in a similar way to NUREG/CR-6909. A key difference between these approaches and the NUREG/CR-6909 is the reduction of conservatisms resulting from the joint implementation of the adjustment sub-factor related to surface finish effect (as quantified in the design air curve derivation) and a F en penalization factor for fatigue assessment of a location subjected to a PWR primary environment. The analysis presented in this paper indicates that the adjustment (sub-)factor on life associated with the effect of surface finish in air (as described in the derivation of the design air curve in NUREG/CR-6909) leads to substantial conservatisms when it is used to predict fatigue lifetimes in PWR environments for rough specimens. The corresponding margins can be explicitly quantified against the design air curve used for EAF assessment, but may also depend on the environmental correction F en factor expression that is used to take environmental effects into account.

Influence of the mean stress on fatigue life and fatigue limit was investigated for carbon steel. Uni-axial fatigue tests were conducted under stress and strain-controlled conditions at room temperature. The fatigue life and fatigue limit were reduced by applying the mean stress for the same stress amplitude. The fatigue life exhibited a better correlation with the strain range than the stress amplitude did. Increase in strain range caused by applying the mean stress correlated well with the decrease in the fatigue life. It was assumed that the mean stress effect on the fatigue life was brought about by the change in crack growth rate caused by applying the mean stress. The mean stress enhanced crack mouth opening and accelerated the crack growth. The reduction in the fatigue limit was also brought about by the same effect. It was shown that the effective strain range gave good prediction of fatigue life and fatigue limit with and without the mean stress.

Requirements for plant license renewal require the evaluation and management of time limited aging analyses for the period of extended operation. These include metal fatigue analyses, with additional consideration for the effects of the reactor water environment. Current license renewal requirements for evaluation of time limited aging analyses for fatigue have extended environmentally assisted fatigue (EAF) considerations to additional locations beyond the originally prescribed NUREG/CR-6260 locations. These requirements have resulted in implementation of a screening process to identify locations that are potentially more limiting for further evaluation of EAF. The screening process includes comparisons of fatigue locations within common bases, including consideration of applicable system transients, rigor of analysis methods, and material types. To date, the United States Nuclear Regulatory Commission has only accepted EAF screening evaluation results based on comparisons within similar material types. However, there are conditions under which a comparison of locations with different material types is valid. This paper presents a basis to compare locations having different material types, to potentially reduce the number of leading (“sentinel”) locations in EAF screening evaluations. The methodology is described through a systematic approach and basis to determine when locations are more limiting with respect to EAF when different material types are compared.

The toughness requirements for the ferritic steels used to construct the primary pressure boundary of a nuclear power plant include both transition temperature metrics as well as upper-shelf metrics. These separate specifications for transition and upper shelf toughness find their origins in decisions made during the 1970s and 1980s, a time when there was much less empirical and theoretical knowledge concerning the relationship between these quantities. Currently, significant evidence exists to demonstrate a systematic relationship between transition and upper shelf toughness metrics for RPV-grade steels and weldments (e.g., the equations in draft Code Case N-830-1, empirical correlation between Charpy transition temperature and upper shelf metrics, etc.). This paper explores these relationships and demonstrates that, in many cases, the joint specification of transition temperature and upper shelf toughness values is redundant and, therefore, unnecessary.

Layered pressure vessels provide a cost-effective solution for high pressure gas storage. Several types of designs and constructions of layered pressure vessels are included in ASME BPV Section VIII Division 1, Division 2 and Division 3. Compared with conventional pressure vessels, there are two unique features in layered construction that may affect the structural integrity of the layered vessels especially in cyclic service: (1) Gaps may exist between the layers due to fabrication tolerances and an excessive gap height introduces additional stresses in the shell that need to be considered in design. The ASME Codes provide rules on the maximum permissible number and size of these gaps. The fatigue life of the vessel may be governed by the gap height due to the additional bending stress. The rules on gap height requirements have been updated recently in Section VIII Division 2. (2) ASME code rules require vent holes in the layers to detect leaks from inner shell and to prevent pressure buildup between the layers. The fatigue life may be limited by the presence of stress concentration at vent holes. This paper reviews the background of the recent code update and presents the technical basis of the fatigue design and maximum permissible gap height calculations. Discussions are made in design and fabrication to improve the fatigue life of layered pressure vessels in cyclic service.

Environmental effects of LWR coolant need to be factored in when defining cumulative fatigue usage of primary circuit components. The basis is a set of codified design rules and fatigue design curves, based on experimental data. To accurately quantify environmental effects, the reference curve in air to which fatigue life in water is compared shall be as reliable as possible. Literature studies and accumulated data at VTT reveal that the use of common reference curves for a wide range of austenitic stainless steel alloys and temperatures is unreliable. Some design codes already include measures to consider this but ASME III is not yet among them. The ASME III design curve is adopted from NUREG/CR-6909 and contains no consideration for dependence of temperature or stainless steel grade. Two different stainless steel grades, AISI 304L and 347, have previously been used in environmentally-assisted fatigue experiments at VTT. In this paper, reference curves for the AISI 304L heat are presented at room temperature and 325 °C to complement the curves already available for AISI 347. Demonstration of realistic environmental effect quantification is done using these reference curves as an alternative to the NUREG methodology.

The primary water environment in a Pressurised Water Reactor (PWR) can have a significantly detrimental effect on the Fatigue Crack Growth (FCG) rates of Austenitic Stainless Steels. Expressions to describe FCG in these materials are provided in ASME Code Case N-809, which was based on results from tests performed under isothermal, simple waveform loading. A previous re-analysis of a much larger database of FCG results highlighted improvements to the N-809 model for nominally low carbon material grades. For non-isothermal and/or complex mechanical loading conditions, further improvements of the prediction of FCG rates were demonstrated by using the Weighted Temperature and K-Rate (WTKR) method. The combined use of improved FCG expressions and the WTKR method is expected to provide significant reductions in over-conservatism when used to assess plant-realistic loading transients. This paper describes the further development of revised expressions to describe the effect of PWR environments on FCG in austenitic stainless steels. The analysis is extended to a wider range of different types of stainless steel, including nominally high carbon 304 variants and 316-type materials. The analysis highlights that previously specified differences in FCG behaviour between nominally low and high carbon materials are minimal, and that 316-type materials exhibit improved performance in these environments. Further testing has also been performed using non-isothermal and complex waveform loading conditions, and these results are used as additional validation of the WTKR methodology.

To ensure an appropriate and/or conservative assessment of structural integrity it is essential to account for the uncertainties inherent to the various inputs and models that, collectively, contribute to a structural integrity assessment. While the methods used to account for uncertainties will differ, this applies equally to assessments performed using either deterministic or probabilistic approaches. Oftentimes the overall model used for a structural integrity assessment is itself comprised of multiple inputs and models, which themselves may be inter-related and/or correlated. In these circumstances the quest to ensure that all uncertainties are addressed can result in the same uncertainty — or uncertainty source — being accounted for multiple times. Such “double-counting” of uncertainties introduces un-needed conservatism to the assessment and should be avoided. In this paper we use the linked fracture toughness models contained in the recently proposed Revision 1 to ASME Section XI Code Case N-830 to provide examples of uncertainty treatment in analyses using multiple models. Identification of sources of uncertainty in each model used in a multi-model analysis can help to ensure that each source is accounted for appropriately and not multiple times. The CC N-830-1 models are used to demonstrate the effects of various uncertainty treatment strategies and the pitfalls that arise from treating sources of uncertainty twice.

High temperature nuclear reactors plan to use highly corrosive coolant such as molten salts, molten lead, and lead-bismuth eutectic mixtures. The existing Class A metallic materials qualified in the ASME Section III, Division 5 rules for high temperature nuclear reactors are not ideal for resisting corrosion when exposed to these coolants. One option to overcome this limitation would be to Code-qualify new corrosion-resistant materials for Class A service, however this process is long and expensive and requires long-term creep test data. A near-term alternative would be to allow designers to clad the existing Class A base materials with non-qualified corrosion-resistant materials. However, there are currently no ASME design rules for cladded components to guard against creepfatigue failure and ratcheting strain accumulation in elevated temperature nuclear service. This work addresses this deficiency by proposing a design strategy for cladded components that does not require long-term testing of clad materials. The proposed approach relies on approximate design analysis methods for two types of clad materials — soft clad that creeps faster than the base material and hard clad that creeps slower and has higher yield stress than the base material. The proposed approach treats a soft clad material as perfectly compliant and a hard clad material as linear elastic. Sample finite element analyses of representative high temperature reactor components are performed to verify the approach. At the end, a complete set of design rules is provided for each of the two types of cladded components.

Fatigue crack growth laws are typically dependent on the ratio between minimum and maximum Stress Intensity Factor (SIF), referred to as the load ratio (R). When part of the SIF range is compressive (and hence R is negative) the amount of growth for a given SIF range is reduced due to crack closure effects. Methods for capturing the effect of crack closure were presented in a previous PVP paper [1]. These methods are based around defining a scaling factor (q 0 ) which is dependent on R and applied to the SIF range before calculated growth. Equations were provided for both best fit and bounding q 0 factors. This paper presents a comparison between these methods and results of testing. The specimens used were square cross-section bars and were made from Type 304L stainless steel with an initial corner crack. A range of load ranges and R ratios (including some positive R values) were used and the testing was undertaken at 250°C in both air and a simulated PWR environment. The growth rate observed in the tests was used to derive the effective q 0 factor observed in each stage of the testing. These values were then compared with the q 0 methods that are used in actual defect tolerance calculations. The results agreed very closely with the derived best estimate q 0 curves, with no discernible difference between the air and water results.

The inspection method based on risk was studied about 23 large tanks of a national strategic oil reserve base, which was a relatively large research work, because the tank inspection was a non-statutory inspection that received little attention, and the risk assessment method of large oil storage tanks in the actual project application was not mature enough. The basic information of 23 nominal volume one hundred thousand cubic meters crude oil tank was collected and sorted. 23 storage tanks under the application of risk assessment are all aboveground atmospheric storage tanks. 23 storage tanks are all floating-roof tanks, whose host materials include SPV490Q, 16MnR, Q235-B, and Q235-A; the containing medium for storage tank are crude oil, and time-to-use starts from August 2006 to April 2007. The implementation process of tank risk assessment was described, and the process of implementing risk assessment was described in detail. The method of risk trend analysis is studied, and the routine maintenance and maintenance suggestions are given. The results show that the risk assessment method for the tank can be implemented, and has been recognized by the tank management unit, providing a rare real case.

In this study, we performed fracture toughness characterization of ten neutron-irradiated Eurofer97 variants using precracked miniature multi-notch bend bar (M4CVN) specimens based on the Master Curve method in the ASTM E1921 standard. The neutron irradiation was performed in the flux trap position of the High Flux Isotope Reactor (HFIR) of the Oak Ridge National Laboratory (ORNL) with the nominal irradiation temperature of 300°C and irradiation dose of 2.5 displacements per atom (dpa). Depending on the irradiation temperature and materials, we observed different degrees of irradiation hardening and embrittlement for ten Eurofer97 variants. The upper shift in the Master Curve reference temperature T 0Q vs. the increase in Vickers microhardness values showed a liner relationship for only a few materials indicating different irradiation responses of the Eurofer97 variants.

The current primary load design provisions of Section III, Division 5 of the ASME Boiler and Pressure Vessel Code, covering high temperature nuclear components, represent an allowable stress methodology using elastic analysis and stress classification procedures to approximate stress redistribution caused by creep and plasticity. This process is difficult to implement and automate in modern finite element frameworks. This paper describes an alternate primary load design approach that uses elastic perfectly-plastic analysis in conjunction with the reference stress concept to eliminate stress classification while retaining a link to the existing Section III, Division 5 allowable stresses. This global, structural allowable stress check is supplemented with a local check to guard against the initiation of creep damage at local stress discontinuities like headers, nozzles, and other stress concentrations. This check is based on a simple elastic-creep analysis with creep damage calculated with the time-fraction approach, using the current ASME minimum-stress-to-rupture values already provided in the current Code. Both the global and local checks are easily implemented in modern finite element analysis software and greatly simplify Section III, Division 5 primary load design when compared to the current design-by-elastic-analysis method. Several examples demonstrate the utility of the new approach and its potential to reduce over-conservatism.

Gas storage well (GSW) is a sort of pressure vessel and a typical model of underground pressure vessel. Now the media of GSW is mainly compressed natural gas (CNG). The main advantages of GSW are analyzed, such as safety and economy are discussed. The current application field and the potential application field of GSW are also explained. Since the idea of GSW originated from oil-gas engineering, GSW has the absolute merit as well as connatural deficiency comparing to the conventional pressure vessel. The problems on material, design, manufacture, inspection, etc. of GSW based on current situation are discussed, forward analysis and expectation for its prospect are brought. With the idea of GSW, some special type underground pressure vessel appeared, such as the underground cylinder type pressure vessel. The structure, design, manufacture of the underground cylinder type pressure vessel is discussed. With the development of new energy, the application field of the GSW is gradually expanded. The hydrogen storage GSW (HGSW) comes into being. The advantage such as safety and economy of HGSW, and the disadvantage are also analyzed. The key questions for the HGSW used in the hydrogenation fueling station are also summarized. The solution for material-hydrogen compatibility and tightness of high pressure HGSW is discussed finally.

Based on the identification of major hazards of pressurized special equipment (PSE) in chemical industry parks, a domain model of major hazards using FCBPSS (Function-Context-Behavior-Principle-State-Structure) method is built in this paper. The characteristics of this model is that it studies major hazards of PSE in the perspective of system by learning its several important concepts, namely function, context, behavior, principle, state and structure, which provides theoretical basis for further study on the dynamic supervision. Furthermore, the model gives powerful technical support to the accident emergency response of pressurized special equipment in chemical industry parks.

Effective stress measures are utilized by engineers to correlate uniaxial and biaxial experimental rupture data to creep failure in real components. Different effective stress formulations are used in various high temperature design codes, such as the Section III Division 5 of the ASME Boiler and Pressure Vessel code and the French RCC-MRx, and in different fitness for service provisions, such as the British R5 and the American FFS-1. Under uniaxial and biaxial stress states all the investigate metrics predict similar rupture times and provide similar safety margins, however there is no guarantee that when these metrics are extrapolated to predict the creep life for triaxial state they will remain conservative. In this work we use the real geometry of a Grade 91 end-plate that failed prematurely and finite element simulations to investigate the variability in creep life predictions obtained utilizing different effective stress as creep-damage driving force. Results show that for triaxial stress states the time to crack nucleation, the crack initiation site and the crack path are extremely sensitive to the effective stress measure.

The typical operating temperatures of a nuclear reactor pressure vessel in a PWR is typically between ∼290°C and 300°C. However, many BWRs and some PWRs operate at slightly lower temperatures down to 275°C. Most of the literature and neutron irradiation damage is therefore focused on those irradiation temperatures. It is well-known that the lower the irradiation temperature, the more neutron irradiation damage occurs, because no appreciable annealing occurs at approximately 230°C. The NOMAD-0 irradiation campaign at the BR2 was a preliminary irradiation specifically designed to determine the appropriate irradiation conditions that result in specific irradiation damage levels of an A508 Cl.2 grade at lower temperatures than the usual PWR operating temperature window. The BAMI capsules with controlled He gas gap were used for this irradiation. To avoid temperature gradients from the outside to the center of the cylindrical blocks, the latter were limited in size. Only tensile and fracture toughness data could therefore be obtained with mini-tensile and mini-CT specimens. The results show that no influence of the temperature gradient could be found on the tensile and fracture toughness properties. The specimen size and geometry on the fracture toughness results was analyzed and discussed. The effect of irradiation temperature on the tensile properties and the transition temperature as determined by the master curve approach on the mini-CT samples is discussed.

Regular inspection is important to verify the integrity of in-service pipeline in the petrochemical industry. Early detection of flaws is therefore essential to ensure the continued safe operation of pipeline. In generally, Conventional ultrasonic for thin-walled small-diameter pipes has limitations due to regulation requirements, efficiencies, lack of permanent results and repeatability. In many cases, the preferred method of inspection is radiographic testing, however there are many obvious shortcomings for radiographic detection. The aim of this work was to propose appropriate phased array ultrasonic scan plans that were able to achieve the following: reduce the inspection times, improve defect detectability and sizing, and reduce human intervention, which will reduce workforce radiation uptake. Ultrasonic simulation using CIVA and experimental verification for thin-walled small-diameter pipe with flaws that were carried out to determine the most appropriate phased array probe and its detection capabilities in this paper.

Gas storage well (GSW) is a type of underground pressure vessel. A series of fatigue and burst tests have been carried out on the simplified gas storage wells with/without thread defects. It can be concluded from the test results that: 1) the quality of the threads and the connecting torque of the threads are the key factor influencing the sealing performance and fatigue property of gas storage wells, the technical requirement of which should be described in the design documents. 2) Under the limited pressure, the threads failure first. The failure mode is pulling-out or leakage. 3) The positions such as connection of well pipe to coupling, connection of well pipe to well head equipment, connection of coupling to well bottom equipment, are the weak parts of the gas storage well. More attention should be paid on these parts in time of designing and in service inspection. 4) for the GSW with only axial thread defect (defect depth < thread height, defect length < thread axial length), the fatigue cycles can be higher than 30,000 and the ultimate strength can be basically equal to the one of the GSW without defects.