Over the past decade, many generation III pressurized water reactor power plants have been under construction in China. Most reactor pressure vessel steels for these plants construction are homemade. Historically, Charpy V-notch specimens are predominantly used to monitor the toughness of RPV steels. However, fracture toughness provides the quantitative predictions of the critical crack size and the allowable stress in structural integrity assessment. This paper evaluates the fracture toughness properties of China manufactured RPV steels directly measured in transition temperature range by using master curve method. Some specimens were irradiated in the High Flux Engineering Test Reactor. The influences of loading rate, test temperature, specimen configuration and neutron irradiation on T 0 were also investigated. The experimental results show that China manufactured RPV steels exhibit good fracture toughness properties.

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.

Brittle fractures in parent material carbon steel pipe, fittings, and flanges are surfacing in recent ASME B31.3 refinery and gas plant construction and facility start-ups with unexpected low toughness of 3J (2.2 ft-lb) to 7J (5.2 ft-lb) at −10°C (14°F) to −29°C (−20°F). The issue is becoming wide-spread globally, affecting up to 30 percent of materials tested, although many manufacturers are not experiencing this issue. The issue creates a new brittle fracture risk that needs to be addressed as the uncertainty of not knowing suitability for service at temperatures down to −29°C (−20°F) is concerning for reliability and safety. These components are considered by ASME VIII Div I and ASME B31.3 Code as being inherently ductile, and brittle fracture resistant without any Charpy impact testing requirements. Testing showed brittle transgranular cleavage cracks. The components were deemed to be unsuitable and not safe for use at low temperatures even though they complied with the applicable ASME Codes [1, 2] and ASTM material standards. Low toughness can result in brittle fracture of the material during hydrostatic tests, cold start-ups, or upset conditions that result in low temperature operations. Additionally, some ASTM A350 LF2 CL1 [3] forged flanges certified to −46°C (−50°F) exhibited the same 3J (2.2 ft-lb) to 7J (5.2 ft-lb) at −46°C (−50°F). This paper discusses historical literature, metallurgical investigations, findings, and factors that contribute to susceptibility to brittle fracture including chemistry, grain size, heat treatment and forming techniques and also issues of ductile to brittle temperature transition shift, and fracture mechanical assumptions. This paper provides guidance to ensure the components are suitable for service and proposes options in addition to the current minimum Codes requirements to mitigate risks of in-service brittle fracture.

UK very high integrity (VHI) component classification includes design, manufacturing, and inspection requirements that go beyond those established in ASME BPVC Sec. III Subsection NB [1]. One of these requirements is to ensure the component is tolerant of manufacturing defects. This can be demonstrated using a Defect Tolerance Assessment (DTA) based on two parameters fracture mechanics method. The brittle fracture parameter of this assessment requires the analysis of stress occurring in the component against the plane strain fracture toughness, K IC of the material. This work focuses on the practical determination of K IC for materials chosen for a Boiling Water Reactor (BWR) Main Steam Piping (MSP) and Main Steam Isolation Valve (MSIV), which carbon steel seamless pipe SA-106 Grade C and carbon steel casting SA-216 Grade WCB, are respectively. These materials are usually tested by Charpy impact testing specified in [1], but there are not many studies reporting their K IC , and there is not enough information concerning actual piping and valve materials. Thus the authors implemented fracture toughness testing using J-resistance curve according to ASTM E 1820 [2] for test pipe and test casting block simulating actual MS Piping and MSIV, and evaluated K IC (J) to be used in DTA. K IC (J) is evaluated from elastic-plastic fracture toughness, J IC , gained from the J-resistance curve, and equivalent to K IC [3]. K IC (J) corresponds to K JIc in ASTM E 1820. There were some cases, however, in which valid J IC values could not obtained, because of the materials high toughness, test specimen size limitations, and uneven final crack sizes. When valid J IC can’t be obtained, retesting or remanufacturing would significantly affect plant construction schedule. Hence, alternative evaluation methods by which J IC can certainly be obtained are desired. In this study, the authors focused on two types of alternative J IC evaluation methods. The first one is the Stretch Zone Width (SZW) method, in which J IC is calculated from SZW measurements of crack tip plastic blunting on fracture toughness test specimens. The SZW method was well studied in the 1970s, and experimental data showed a clear correlation between J IC values obtained from J-resistance curves and J IC values obtained from SZW measurements [4]. The second method is by correlation of J IC with the energy absorbed during Charpy testing. As represented by Rolf’s study [5], it has been reported that there are correlations between Charpy absorbed energy and K IC for high tensile strength steels. In this study, the validity of the SZW method was first evaluated by comparing its results with J IC obtained from J-resistance curves. Then, the applicability of the J IC values to DTA of actual products was discussed. Finally, by comparing Charpy absorbed energy and K IC (J) , the validity and applicability of K IC determination method with Charpy absorbed energy was discussed.

Authors: Sargent & Lundy, LLC: Delfo Bianchini, Dennis Demoss and William Peebles. Focus: Operations, Applications & Components. This paper will review new Small Modular Reactor (SMR) designs and implications on the ASME Section III Code. Four SMR technologies will be discussed: B&W mPower (TVA Clinch River site), Westinghouse SMR (Ameren Callaway site), NuScale/Fluor SMR (Pacific NW lab site) and Holtec SMR (Savannah River site). SMR reactor designs vary from 45 to approximately 225 MWe. SMR goals include significantly reducing plant capital cost requirements and enabling multi-reactor module construction and addition over time providing greater utility flexibility. A primary SMR advantage includes its installation in smaller grids typical of electrical power systems in developing countries. Unique aspects of the SMR technologies include integral reactor and steam generator vessel, integral pressurizer and internal piping, below grade containment vessel, helical-coil integral steam generators, integral decay heat removal systems, modular plant construction and arrangement, in-service inspection unique requirements, special materials and welding. Current ASME code design rules and requirements may require expansion and clarification; Section XI testing requirements and frequencies may require revision.

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The US NRC developed leak-before-break (LBB) procedures in draft Standard Review Plan (SRP) 3.6.3 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987 with a modest revision in 2005. The United States Nuclear Regulatory Commission (US NRC) is currently developing a draft Regulatory Guide on what is called the Transition Break Size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1e−6 per year, that a Double-Ended Guillotine Break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94-percent of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1e−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.

The Atucha II nuclear power plant is a unique pressurized heavy water reactor being constructed in Argentina. The original plant design was by KWU in the 1970’s using the then German methodology of break preclusion, which assumed that the largest break-opening area would be 10-percent of the cross-sectional area of the largest pipe diameter. That philosophy was used for the design of the emergency core cooling system in the 1970’s. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The construction is progressing with initial start-up in 2011. Since the 10-percent of the cross-sectional area is a smaller ECCS design requirement than the normally assumed double-ended-guillotine break, the safety evaluation of the plant for beyond design basis seismic loading of the nuclear plant was a regulatory requirement. This overview paper describes a Robust LBB Evaluation that was conducted in great detail to assess the safety aspects of the piping system under beyond design basis seismic loading and the implications to the ECCS. Key aspects involved: • Static and dynamic material property testing, • Determination of weld residual stresses, • Determination of crack sizes that might evolve by worst case SCC growth rates under weld residual stresses and normal operating stresses, • Determination of leakage rates as a function of time with the upper-bounding crack growth rates, • Development of seismic hazard curves for the site, • Development of FE models of the containment building and primary NSSS system within the building, • Determination of normal operating stresses, SSE stresses and 10 −6 seismic stresses using worst case soil foundation assumptions, • Evaluation of flaw behavior for circumferential cracks using the shapes from the natural crack growth. • Evaluation of margins on the critical flaw size and times to leakage, and • Standard LBB analyses, as well as Transition Break Size evaluations. The key result from this effort was that even with all the normal operating plus 10 −6 seismic event loading, the pipe system behaved more like it was displacement-controlled than load-controlled. The displacement-controlled behavior made the pipe much more flaw tolerant, and it was found that a DEGB was not possible because the flaw could never reach the critical flaw size without greatly surpassing the leakage and water make-up capacity of the plant. Since there are many details in this multi-year effort, only the key points will be summarized in this paper while other details will be the topics of other papers.

Hot Isostatic Pressing (HIP) has been used for many years to consolidate porosity in cast metal shapes to improve mechanical properties. When applied to fine metal powders, it is possible to produce Near Net Shape (NNS) items and more complex geometry components that are fully dense and offer an attractive set of properties at reduced cost. NNS items produced from HIPed powder deliver cost savings by reducing initial material usage and subsequent machining costs. Powder production and HIP processing are automated methods, which provide protection against forging route obsolescence. Setup costs are lower and batch sizes are smaller, which makes HIPping particularly well suited to small numbers of high integrity components. HIPed powder microstructures are isotropic and equiaxed, with uniformly fine grain sizes not normally achieved in heavy section components, which facilitates ultrasonic NDE examination. Improved features to facilitate NDE are readily incorporated into the HIP assembly. Inclusion contents are lower and of more benign geometry, easing fracture assessment. In a broad program of testing, Rolls-Royce has established (1) that HIPed powder 316L/304L components, in items up to several tons in weight, have equivalent or slightly better strength, toughness and corrosion resistance than the wrought equivalents. Rolls-Royce are extending their activities to HIPing of Inconel alloys. The first phase has been to HIP test samples of Inconel 600 and Inconel 690 alloys. Initial testing has produced promising results in line with expectations of wrought material. There has also been the opportunity to vary the HIPing cycle to assess the effect of processing parameters on the final product. An ability to HIP Inconel components is thought to be of benefit in new plant construction, where material is often not readily available in required thick section. The adaptability and good control of the HIP technique also shows promise as a manufacturing route for future high temperature materials which will be required in Generation 4 civil builds.

Hot Isostatic Pressing (HIP) has been used since the 1980s to consolidate porosity in cast metal shapes and improve mechanical properties in conventional forgings and wrought components. The availability of high quality metal powders has made it possible to produce Near Net Shape (NNS) items and more complex geometry components that are fully dense and offer an attractive set of properties at reduced cost. Powder HIP manufacturing reduces initial material usage and subsequent machining costs. Metal powder production and HIP processing are automated methods, which also provide protection against forging route obsolescence. Setup costs are lower and batch sizes smaller. HIPped powder microstructures are isotropic and equiaxed, with fine grain sizes not normally achieved in heavy section components, which facilitates ultrasonic NDE examination. Inclusion contents are lower and of more benign geometry, easing fracture and safety case development. Although widely used in the off-shore oil industry in high integrity applications, particularly to reduce welded connections, in the nuclear industry interest has been limited. The quality of HIPped powder items can provide through life cost savings since there is greater assurance of structural integrity compared to welded or wrought components. In an extensive programme of testing, it was established that HIPped powder 316L and 304L components, in items up to several tons in weight, have equivalent or slightly better strength, toughness and corrosion resistance. HIPped powder items are now in service as pressure retaining components in PWR plant. Effort is now directed at widening the range of components for which the HIP process is appropriate focusing on reducing welds in the plant construction sequence. This is particularly relevant to pipework manufacture and assembly. The benefits of facilitating an ASME Code Case for Powder HIP are also being considered.

In order to introduce environmental effects into fatigue evaluation of design and construction codes, the environmental fatigue evaluation method should not only be established, but the conservativeness of the codes, such as safety factors of design fatigue curve and simplified elastic-plastic analysis method ( K e factor), etc. should also be reviewed. Then plant designers should optimize total system of fatigue evaluation according to the objective of the codes by properly selecting design transient conditions and stress analysis methods used in fatigue evaluations as necessary. In addition, investigation of measures for reducing fatigue should be performed to mitigate possible fatigue initiators and alternative evaluation methods in case that the evaluation result should exceed the criteria specified in the design and construction codes. This paper discusses the present status in the review of these items for the Japanese PWR plants and future prospects to tackle on the application of environmental fatigue evaluation in design stage of plant construction.