May-June 2019 NPJ

16 NuclearPlantJournal.com Nuclear Plant Journal, May-June 2019 New Documents EPRI 1. Technical Basis for Optimization of the Volumetric Examination Frequency for Reactor Vessel Studs . Product ID: 3002014589. Published December, 2018. Rules for periodic inspections of nuclear power plant components are provided in Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Within Section XI, the Class 1 components requiring periodic inspection during each code interval are included in Table IWB-2500-1. Among these, Examination Category B-G-1, Pressure-Retaining Bolting, Greater Than 2 in. (50 mm) in Diameter, Reactor Vessel Item No. B6.20, “Closure Studs” requires periodic volumetric or surface examination of all reactor pressure vessel (RPV) closure studs every Inspection Interval (nominally 10 calendar years). This report develops a technical basis for optimizing the frequency of Item No. B6.20 examinations. The technical basis considers the primary degradation mechanisms applicable to RPV studs, including (1) fatigue, (2) stress corrosion cracking, (3) boric acid corrosion (pressurized water reactors only), and (4) steam cutting. Although the technical basis is oriented toward ASME Code Section XI requirements, the analysis approach and results have merit as a stand-alone technical position. International utilities that use different governing codes and standards for inspections should evaluate how to use the report in conjunction with those standards and regulatory obligations. Given the operating experience to date for RPV studs, the quantitative assessments in the technical basis report focus on the potential for RPV stud degradation caused by fatigue mechanisms. The technical basis in the report for the optimization of Item No. B6.20 inspections for RPV studs includes (1) assessing typical design basis loads and transients, (2) evaluating the stresses using a finite element analysis of the reactor vessel head closure, (3) identifying and evaluating flaw stability limits, and (4) evaluating fatigue crack growth of a postulated flaw in the RPV studs. The time for the postulated flaw to propagate beyond an acceptable flaw size can be used to optimize an appropriate inspection frequency. 2. Technical Update: Technical Basis for Selection of Mitigation App . Product ID: 3002016788. Published May, 2019. The detonation of a nuclear weapon at high altitude or in space (~ 30 km or more above the earth’s surface) can generate an intense electromagnetic pulse (EMP) referred to as a high-altitude EMP or HEMP. The E1 EMP component of HEMP can propagate to the earth and impact various land-based technological systems such as the electric power grid. Of particular concern is the potential impact of E1 EMP on critical electronics- based assets such as digital protective relays and communications equipment. The Electric Power Research Institute (EPRI) launched a three-year research project in April 2016 to investigate the potential impacts of a HEMP attack on the electric transmission system and to identify possible options for mitigating impacts. As a part of this research effort, testing was performed to investigate how DPRs and communications equipment within a substation control house would perform when subjected to both radiated E-fields and conducted surges. In addition, the levels of shielding provided by typical substation control house designs, which may reduce the E-field magnitudes incident on the DPRs, were evaluated, as well as the performance of surge protection devices. This report documents the results of testing that determines the incident radiated E-fields and the surge voltages that the digital protection relays (DPRs) and communications equipment that was tested can withstand so that the vulnerability can be determined. In addition, this report presents the results of testing mitigation methods, including shielding of substation control houses and surge protection. 3. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to- Shell Welds and Nozzle Inside Radius Sections . Product ID: 3002014590. Published April, 2019. Certain welds in pressurized water reactor (PWR) steam generators (SGs) are classified as Class 2, Category C-B, pressure retaining welds in pressure vessels. This report focuses on the nozzle- to-shell welds and the inside radius sections of PWR SG feedwater and main steam nozzles, which are listed in ASME Code Section XI, Table IWC-2500-1 under the following item numbers:  Item No. C2.21: Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) weld.  Item No. C2.32: Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds when inside of vessel is accessible.  Item No. C2.22: Nozzle inside radius section. The requirements for Item No. C2.21 call for both surface and volumetric examinations, while Item Nos. C2.22 and C2.32 require only volumetric examination. These items require examination of all nozzles at the terminal ends of piping runs during each inspection interval, which means that all nozzle- to-shell welds and nozzle inside radius sections are examined every interval. The objectives of this report are to evaluate the current examination requirements for PWR SG feedwater and main steam nozzle-to-shell welds and nozzle inside radius sections and establish the technical bases for various alternative inspection scenarios. To accomplish these objectives, various topics are addressed in this report, including a review of previous related projects, a review of inspection history and results, a survey of components in the industry, selection of representative components and operating transients for stress analysis, evaluation of potential degradation mechanisms, and a flaw tolerance evaluation consisting of probabilistic and deterministic fracture mechanics analyses. 4. Nondestructive Evaluation Program Highlights: 2019 Vol. 2, March/April . Product ID: 3002016132. Published April, 2019. This newsletter is published monthly by EPRI to report on activities related to nondestructive evaluation (NDE). The above EPRI documents may be ordered by contacting the Order and Conference Center at (800) 313-3774, Option 2, or email at orders@epri.com .