Reliability Assessment of Piping With Cracks Using Advanced First-Order Second-Moment Method

This paper describes applicability of the 2 parameter assessment method using a reference stress method from the viewpoint of reliability. The applicability of the reference stress method was examined comparing both the GE-EPRI method. As a result, J-integral and limit load at the time of fracture evaluated by the reference stress method is almost equivalent to that by the GE-EPRI method. Furthermore, the partial safety factor (PSF) evaluated by reliability assessment has little difference between two methods, and the required safety factor is enveloped by the safety factor for Service Level-A and B defined in fitness for service (FFS) codes. These results show that of the reference stress method is applicable for J-integral calculation in fracture assessment.

Numerical Simulation on the Sound Field of Gas Valve Due to Inner Leakage

Recently, acoustic testing method has been widely used for valve leakage inspection. Traditional method has difficulty carrying out research on the acoustic testing of valve leakage due to the complex geometry of the valve and the effect of flow on sound field. Thus, a numerical method was put forward based on Reynolds average N-S equation and k – ε equation in order to investigate the flow of the inner leakage of ball valve, needle valve and gate valve. Furthermore, difference time domain method was adopted to simulate the noise field of the flow based on Lighthill acoustic equation. And the characteristics of acoustic distribution were obtained. The results showed that turbulent flow was the main cause of jet noise inner the valve due to gas leakage. And the acoustic distribution had directivity. There were about 10 ∼ 25dB difference between the upstream and downstream of the leaking valve. The position with the highest sound pressure was also found out.

Effect of Additional Irradiation at Different Fluxes on RPV Embrittlement

The effect of the neutron flux at high fluence on the microstructural and hardness changes of a reactor pressure vessel (RPV) steel was investigated. An accelerated test reactor irradiation of a RPV material, previously irradiated in commercial reactors, was carried out at the lowest possible neutron fluxes in order to obtain neutron fluences up to approximately 1×1020 n/cm2 (E>1MeV). State-of-the-art experimental techniques such as three-dimensional atom probe were applied to carry out advanced quantitative characterization of defect features in the materials. Results for the same material irradiated in both high and low flux conditions are compared. For neutron fluences above 6×1019 n/cm2 (E>1MeV) the difference in the neutron fluence dependence of the increase in hardness is not seen for any neutron flux condition. The volume fraction of solute atom clusters increases linearly with neutron fluence, and the influence of neutron flux is not significant. The component elements and the chemical composition of the solute atom clusters formed by the irradiation do not change regardless of the neutron fluence and flux. The square root of the volume fraction of the solute atom clusters is a good correlation with the increase in hardness.

BS 7910: History and Future Developments

BS 7910, the UK procedure for the assessment of flaws in metallic structures, was first published almost 30 years ago in the form of a fracture/fatigue assessment procedure, PD6493. It provided the basis for analysing fabrication flaws and the need for repair in a rational fashion, rather than relying on long-established (and essentially arbitrary) workmanship rules. The UK offshore industry in particular embraced this new approach to flaw assessment, which is now widely recognised by safety authorities and specifically referred to in certain design codes, including codes for pressure equipment. Since its first publication in 1980, PD6493/BS 7910 has been regularly maintained and expanded, taking in elements of other publications such as the UK power industry’s fracture assessment procedure R6 (in particular the Failure Assessment Diagram approach), the creep assessment procedure PD6539 and the gas transmission industry’s approach to assessment of locally thinned areas in pipelines. The FITNET European thematic network, run between 2002 and 2006, has further advanced the state of the art, bringing in assessment methods from SINTAP (an earlier European research project), R6, R5 and elsewhere. In particular, the FITNET fracture assessment methods represent considerable advances over the current BS 7910 methods; for example, weld strength mismatch can be explicitly analysed by using FITNET Option 2, and crack tip constraint through Option 5. Corrosion assessment methods in FITNET are also more versatile than those of BS 7910, and now include methods for vessels and elbows as well as for pipelines. In view of these recent advances, the BS 7910 committee has decided to incorporate many elements of the FITNET procedure into the next edition of BS 7910, to be published c2012. This paper summarises the history of the development of BS 7910, its relationship with other flaw assessment procedures (in particular FITNET and R6) and its future.

Experimental Study of Ductile Fracture for Non-Aligned Multiple Flaws in a Plate

The fitness-for-service code requires the characterization of non-aligned multiple flaws for the flaw evaluation, which is performed using a flaw proximity rule. Worldwide almost all codes provide own proximity rule, often with unclear technical bases of the application of proximity rule to ductile fracture. To clarify the appropriate proximity rule for non-aligned multiple flaws in fully plastic fracture, fracture tests on flat plate specimen with non-aligned multiple through wall flaws were conducted at ambient temperature. The emphasis of this study was put on the flaw alignment rule, which determines whether non-aligned flaws are treated as independent or aligned onto the same plane for the purpose of flaw evaluations. The effects of the flaw separation and flaw size on the maximum load were investigated. The experimental results were compared with the estimations of the collapse load using the alignment rules in the ASME Section XI, BS7910 and API 579-1 codes. A new estimation procedure specific to the fully plastic fracture was proposed and compared with the comparison with the experimental results.

Evaluation of Alignment Rules Using Stainless Steel Pipes With Non-Aligned Flaws

Multiple flaws such as stress corrosion cracks are frequently detected in the same welded lines in pipes. If multiple discrete flaws are in close proximity to one another, alignment rules are used to determine whether the flaws should be treated as non-aligned or as coplanar. Alignment rules are provided in fitness-for-service codes, such as ASME, JSME, API 579, BS 7910, etc. However, the criteria of the alignment rules are different among these codes. This paper briefly introduces these flaw alignment rules, and four-point bending tests performed on stainless steel pipes with two non-aligned flaws. The experimental plastic collapse stresses are determined from the collapse loads and compared with collapse stresses calculated from the limit load criteria. The limit loads are obtained for single non-aligned or aligned coplanar flaws in accordance with the alignment rules. On this basis, the conservatism of the alignment rules in the above codes is assessed.

Methodology for the Determination of a Set of Safety Factors That are Consistent for Design Code and Fitness-for-Service Code: Framework

This paper introduces a methodology for the determination of a complete set of safety factors that maintains consistency between design code and fitness-for-service code of nuclear components. The purpose of the work is to materialize the System Based Code concept, which is indispensable for the development of next generation nuclear reactors. The methodology consists of three principles proposed by the author which should be the basis of code development for new next generation reactors. The principles are; 1) Design to target reliability, 2) Continuous reliability evaluation from design to fitness-for-service, 3) Update of reliability evaluation based on information obtained during construction and operation. Effectiveness of the methodology is demonstrated using a simple example problem. The problem deals with pipe subjected to internal pressure under conditions which is typical in light water reactors. Following the reliability evaluation of current situation which meets the provisions of design code and fitness-for-serve code published from Japan Society of Mechanical Engineers, the three principles are applied step-by-step and safety factors and reliability indices are newly derived. It is shown that a complete application of the three principles could lead to a set of safety factors that assures consistency in terms of reliability in design and fitness-for-service, and improves allowable stresses as well. Technologies to be developed and issues to be discussed for application of the methodology to more complicated and practical situations are described as well.

Failure Assessment of Nuclear Piping Components Due to Combined Degradation Mechanisms by an Advanced Probabilistic Fracture Mechanics Code

Probabilistic failure analysis of nuclear piping components due to a combined degradation mechanisms is a challenging issue. At present the majority of analyses were done by assuming a single failure mechanism for a specific location of a piping system. But in reality, this might not be an accurate approach. A tiny crack might be present in a weld location due to fabrication defect or initiated due to fatigue after a short incubation time of plant’s start up. This pre-existing or initiated crack then may be further matured by the synergistic effect of different probable degradation mechanisms e.g. fatigue, stress corrosion cracking, etc. In this study the development process of an advanced probabilistic fracture mechanics code has been described which can handle this combined failure mechanisms. Numerical examples are also presented to rationalize the use of such methodology.

Stress Corrosion Cracking in Welds of Reactor Vessel Nozzle at Ohi-3 and of Other Vessel’s Nozzle at Japan’s PWR Plants

During the 13th periodic inspection, which started in February 2008, KEPCO’s Ohi unit 3 (1,180MWe PWR) implemented voluntary ECT in addition to the visual inspection of the RV hot and cold leg nozzle welds to confirm the integrity of the concerned section. As a result of inspection, in March 2008, a flaw extending in the depth direction along dendritic grain boundaries of the weld metal was found in the RV A-loop hot leg nozzle. With traces of machining, which could cause residual tensile stresses, it was suspected that SCC initiated and grew at the concerned section. After grinding the section to remove the entire flaw, WJP was applied as the corrective action. Ohi-3 restarted operation on November 2008. It is planned to apply repair welding to the ground out section with alloy 690 during the next periodic inspection [1]. Several Japan’s PWR plants have experienced similar incidents in the nozzle welds. This paper presents the details of repair technologies, which have been developed to address PWSCC found in Ohi-3 RV hot leg nozzle and previous similar incidents in the RV and other vessel’s nozzle welds at Japan’s PWR plants.

Flaw Tolerance Evaluation of CASS Piping Materials

Prolonged exposure of cast austenitic stainless steels (CASS) to reactor coolant operating temperatures has been shown to lead to some degree of thermal aging embrittlement (reduction in fracture toughness of the material as a function of time). The fracture toughness data for the most severely aged CASS materials were found to be similar to those reported for some austenitic stainless steel weld metal, in particular weld metal from submerged arc welds (SAW). Such similarity offers the possibility for applying periodic inservice inspection flaw acceptance criteria, currently referenced in the ASME Code Section XI, Subsection IWB, for SAW and shielded metal arc weld (SMAW), to CASS component inservice inspection results. This paper presents the data to support both the proposed screening criteria (based on J-R crack growth resistance) for evaluation of the potential significance of the effects of thermal aging embrittlement for Class 1 reactor coolant system and primary pressure boundary CASS components, for those situations where the effects of thermal aging embrittlement are found to be potentially significant. The fitness for continued service is based on the comparison of the limiting fracture toughness data for Type 316 SAW welds and the lower-bound fracture toughness data reported for high-molybdenum, high delta-ferrite, statically and centrifugally-cast CASS materials. These comparisons and the associated flaw acceptance criteria can be used to justify exemptions from current ASME Code Section XI inservice inspection requirements through flaw tolerance evaluation (e.g., see ASME Nuclear Code Case N-481).