Steam dryers in Boiling Water Reactors, located in the upper steam dome of the reactor pressure vessel, are not pressure retaining components and are not designed and constructed to the provisions of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. As such, these components do not correspond to any specific safety class referenced in the Code. Although the steam dryers in BWRs perform no safety function, they must maintain the structural integrity in order to avoid the generation of loose parts that may adversely impact the capability of other plant equipment to perform their safety functions. Therefore guidance from Section III of the ASME Code is utilized in the design and fabrication of replacement dryers as well as for design modifications of the existing dryers for extended power uprates. The majority of licensees of operating nuclear plants are applying for EPU, which generally increases the thermal power output to 20% above the original licensed thermal power. Nuclear power plant components such as steam dryers can be subjected to strong fluctuating loads and can experience unexpected high cycle fatigue due to adverse flow effects while operating at EPU conditions. However, there are some unique challenges related to steam dryer operation at EPU conditions requiring special considerations to prevent fatigue damage from the effects of flow induced vibration. This paper examines the issues and lessons learned related to FIV considerations during EPU reviews of BWR steam dryers.
In 1997, the joint ASME-QME/IEEE-NPEC Special Working Group on Standardization of Experience Based Seismic Qualification developed a recommendation for incorporation of experience-based seismic qualification of equipment into the Qualification of Mechanical Equipment (QME) Standard. In response to this recommendation, the QME Main Committee formed a Subgroup on Dynamic Qualification and chartered this Subgroup to incorporate experience-based seismic qualification of equipment into Appendix A of Section QR of the QME-1 Standard. This paper provides an update on the progress of the ASME QME Subgroup on Dynamic Qualification in developing an update to the QME-1 Standard that will include the use of earthquake experience for the seismic qualification of mechanical equipment used in nuclear facilities.
Due to its recognized resistance against galvanic corrosion, microbiological induced corrosion (MIC), and fouling compared to carbon steel, high density polyethylene (HDPE) piping material has recently been proposed for replacement of carbon steel piping in nuclear safety related class 3 service water buried piping applications. However, there are some unique challenges requiring special design considerations due to HDPE’s visco-elastic nature, need for relatively thicker pipe walls, and very low material allowable stress. This paper examines special design considerations such as design factor, minimum required wall thickness for pressure, thermal gradient effects, and cumulative effect from loads of different duration. Useful guidelines from design and analysis perspective are provided.
Carbon fiber reinforced polymer (CFRP) composites have been used for decades in various industries such as aerospace, oil and gas, and transportation mainly due to their high strength-to-weight-ratio and excellent corrosion resistance. However, the use of CFRP in nuclear industry applications has been very limited. Recently, a new ASME Boiler and Pressure Vessel (BPV) Code Case N-871 has been proposed for internal repair of buried Class 2 and 3 nuclear safety related piping using CFRP for Service Levels A, B, C and D at temperatures not exceeding 200F. This is a first-of-a-kind CFRP application for nuclear safety related piping. It is known that CFRP materials are subject to property degradation due to environmental exposure as well as creep behavior under sustained loading. These factors should be considered when designing the CFRP repair for any nuclear safety related piping application to ensure an adequate safety margin is maintained. In the proposed Code Case, there are provisions for using two different design (analysis method) methods — Allowable Stress Design (ASD) and Load and Resistance Factor Design (LRFD) methods. The LRFD method has been widely used in civil engineering applications but has never been used in ASME Section III Code for nuclear piping applications. This paper presents a comparison of these two methods from a safety margin point of view. As CFRP is a new material for ASME BPV Code for nuclear safety related applications, several areas have been identified in the design concepts to ensure an adequate safety margin for Service Levels A, B, C and D. An effort is also made to provide guidelines on the required safety margin for CFRP repair of safety related piping. Finally, Code Case N-871 is reviewed to evaluate the minimum safety margin available for both ASD and LRFD methods.
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