July-August 2019 NPJ

30 NuclearPlantJournal.com Nuclear Plant Journal, July-August 2019 Reliability & Integrity... ( Continued from page 29) 2. After SCCs for the RIM process are identified, all potential degradation mechanisms (DMs) are considered, including those that are not explicitly addressed in design and construction codes (e.g., flow induced vibration). The concept of the DM assessment is to consider whether any of the following DMs might apply to a specific reactor designs and related components:  Design characteristics, including materials, component type, and other attributes related to the system configuration.  Fabrication practice introduced DMs, including welding and heat treatment.  DMs introduced by operating and transient condition, including: temperatures, pressures, and gas/water flow, fluid quality (e.g., primary water, rawwater, dry steam, chemistry control, etc.), and other service environments (e.g., humidity, radiation, etc.).  DMs based on plant-specific or industry- based service experience, if available.  Results from pre-service, inservice, and augmented examinations and the presence and impact of any prior repairs in the SSC.  Manufacturer’s recommendations for examination, maintenance, repair, and replacement. These DMs can then be modeled and evaluated to the extent that otherwise are not considered in many design codes. For example, crack propagation and resultant fracture could be included in a RIM evaluation if appropriate. 3. Once the DM assessment is completed for an SSC, all credible failure modes are identified based on the associated degradation mechanisms for each SSC being evaluated. 4. When the DM Assessment is completed, the probability of each failure mode is assessed for the SSC. 5. At this stage of the process, required RIM evaluations additionally consider the safety functions of the plant, taking into account events that have been postulated in the safety analysis of the plant and Probabilistic Risk Assessment (PRA), such as:  The plant operating state (e.g., mode or operational condition such as hot standby) relevant to the plant level risk and reliability goals and SSC level reliability targets.  Initiating events including internal events and events associated with internal and external plant hazard.  Event sequence development sufficient to support the quantification of mechanistic source terms and offsite radiological consequences consistent with applicable regulatory limits on the frequencies and consequences of licensing basis events. The main focus of evaluations in RIM is on the probability of occurrence of an event (e.g., to meet plant safety analysis and regulatory criteria). The ability to adequately identify and mitigate an event(s) as defined in a plant’s safety analysis (e.g., detect a pipe break) is also considered in this phase of the RIM evaluation. DM assessment, credible failure modes, and established failure mode probabilities are integrated into a plant’s PRA to establish SCC’s required reliability targets. 6. The aggregated information from all previous evaluations is then used to select meaningful inservice inspection (ISI) methods with appropriate monitoring and inspection frequencies that would then be used in developing specific ISI provisions for any given particular SSC. In this regard, RIM introduces a new concept for ISI known as MANDE (i.e., monitoring and nondestructive examination) and the development of MANDE is assigned to a required expert panel known asMonitoring and NDE Expert Panel (MANDEEP). This panel establishes the appropriate MANDE criteria for each SSC within the program, so the SCC is actively evaluated for the onset of any credible damage mechanism(s) and assures that selected MANDE supports and maintains established SSC’s Reliability Targets throughout its life cycle. The MANDEEP is responsible for overseeing the qualification of MANDE methods and techniques in accordance with Mandatory Appendix IV, Monitoring and NDE Qualification, under the RIM Program. This Appendix provides requirements for performance-based qualification of monitoring and nondestructive examination methods and techniques. The MANDEEP addresses qualification of the personnel, procedures, and equipment. A performance-based system verifies and documents that MANDE methods and techniques that are selected are capable of performing tasks to produce outputs to meet defined industry requirements as shown in Figure 1, MANDEEP Performance-Based Process. 7. Like most inservice inspection programs that are founded on risk insights, it is expected that once a RIM program is developed and implemented, it will be regularly monitored for effectiveness and employ a continuous feedback loop used to update and make adjustments to the program as additional operating experience is obtained or if the parameters used to develop the original RIM application have changed. This integrated process is the foundation of the RIM. It is the over- arching foundation for the development of ASME Code Section XI, Division 2 (RIM), and when employed, will serve to effectively provide high reliability for critical SSC – from initial operations through end of life service by effectively creating a comprehensive ageing management program (AMP) that protects SSC reliability over their entire life cycle. In depth details for this process may be found in Reference 1. Need for ASME Code Section XI, Division 2 Shifts in approach for ISI of advanced reactor designs are necessary to accommodate variables such as refueling cycles, MANDE methods for degradation mechanisms that may not lend themselves well to detect the onset of credible degradation mechanisms (e.g., creep), and unique operating parameters that may apply to future reactor designs (e.g., atmospheric operating pressures) . Under current ASME Code Section XI, Division 1 rules, ISI examinations are prescriptively required to be performed at discrete time periods during a typical 10- year inservice inspection interval. Some advanced reactor designs may be designed for longer fuel cycles than today’s typical PWR or BWR 18-to-24-month fuel cycle. This means that the typical 10-year ISI program interval may not be well suited for these advanced reactor designs. Real