12 Checks When Qualifying Piping Systems in Nuclear Applications

12 Checks When Qualifying Piping Systems in Nuclear Applications


The analysis and qualification of piping systems in nuclear power plants involves more than meeting Code stress limits. Generally, a piping system is qualified if the following criteria have been met. These various qualification criteria are typically specified in the plant FSAR, the plant design procedures, or the ASME Code.

  1. Pressure design in accordance with the design Code. This check will govern the schedule of pipes, the thickness of tubing, the schedule and pressure class of fittings, the reinforcement of openings and branch connections, the pressure rating of valves, the pressure class of flanges, and the pressure design of specialty fittings.
  2. ASME Code stress limits. This check will verify that the Code stress equations for the load combinations corresponding to Test, Design, and Service Levels A, B, C, and D loadings have been met.
  3. NRC Standard Review Plan (SRP) Section 3.6 stress limits for the postulation of high energy line breaks. This check will verify that high energy line breaks have been postulated at locations where the stresses exceed the SRP 3.6 limits. This check will also verify that, for pipes in break exclusion zones (BEZ), the SRP 3.6 stress limits for the BEZ have been met.
  4. Functional capability stress limits. Some plants have committed in Chapter 3 of the FSAR to meet restrictive stress limits for Service Level D load combinations. This was done at the time to assure that the pipes would not deform or buckle to the point of restricting flow, for lines that have to operate following the postulated Service Level D event. While this concern has since been proven to not be technically valid (NUREG-1367, “Functional Capability of Piping Systems”, D. Terao and E. Rodabaugh, 1992), the restrictive stress limits remain a licensing commitment for many plants, and have to be checked.
  5. Stress limits at welded attachments. This check verifies that the stresses at welded attachments (typically cylindrical trunions or rectangular lugs) meet stress limits specified in ASME III Appendix Y. This check is not explicitly called-for in B31.1, but many plants have committed in their FSAR to check welded attachment stresses in safety-related piping systems. This stress check is more than verifying the strength of the fillet welds at the attachments, they are meant to prevent a tear of the weld or the base metal near the weld.
  6. Nozzle load limits. This check will verify that the loads on nozzles (three forces and three moments) are within the limits specified by the manufacturer, and the ASME III Design Specification. Nozzle load limits for static equipment (vessels and tanks) are set by stress limits in the nozzle and the equipment shell. Nozzle load limits for rotating equipment (pumps, compressors) are set by the manufacturer to limit distortion of the equipment which would impede its operation. Some ASME III plants may also have load limits on valve nozzles, although ASME III NC/ND-3521 does not require limits if the section modulus of the valve body crotch is at least 10% thicker than the pipe, and the valve material strength is at least that of the pipe.
  7. Penetration load limits: Wall and containment penetrations are designed within strict load and movement limits. These loads and movements are caused by the penetrating pipe, and also by the relative expansion of the wall and containment under normal environmental temperature fluctuations and large heat-up and pressurization following a postulated pipe break.
  8. Loads and/or movement limits at mechanical joints. ASME III NC/ND-3658 (but not B31.1) has pipe load limits on ASME B16.5 flanges. Manufacturers of mechanical joints (such as expansion joints, grooved couplings, swaged couplings) have load and/or movement limits on their joints.
  9. Support loads and movements. These are the result of the stress analysis and are input to the design of the support. In some plants, the supports are also checked for stiffness by verifying that the support does not deflect more than a specified distance (e.g., 1/16 in.) in the direction of load. In common practice the piping analysis model does not include friction, so the friction force must be added to the support reaction loads. This is often done by adding a friction force in the direction where the support has a thermal expansion movement larger than 1/16 in. For small-bore piping some plants have designed supports for at least a minimum load, regardless of the magnitude of the support load from the stress analysis output. More details on rules of good practice for support design are addressed in WRC Bulletin 353 “Position Paper on Nuclear Plant Pipe Supports”, 1990.
  10. Clearance and rattle points. Some plants include in their piping analysis procedures a check of the calculated pipe movements against measured clearances or rattle points, if the calculated movement exceeds a certain threshold, for example 2 in.
  11. Flow-induced vibration (FIV). The prediction of pressure pulsations in piping systems, either caused by the pump-compressor, or by flow instabilities, vortex shedding, or at the extreme by cavitation, and their potential amplification acoustically or structurally is very difficult to predict. That is why flow-induced vibration is typically not accounted for at the design stage. Instead, FIV is addressed in two ways: (a) Following rules of good practice in hydraulic design and layout, and (b) for a new systems or hydraulic modifications to existing systems, FIV is checked by visual observation during system start-up.
  12. Acceleration limits of valve operators. The operators of active valves (air-operated, motor-operated, solenoid operated) that have to operate following a seismic event, are qualified to sustain a certain acceleration without loss of function. The qualification is typically performed by shake table testing (IEEE-344 or ASME QME-1), with peak seismic accelerations in the range of 3g for OBE and 6g for SSE. The analysis must then verify that the total seismic acceleration at the valve operator does not exceed these limits.

About The Author

George Antaki, Fellow ASME, has over 40 years of experience in nuclear power plants and process facilities, in the areas of design, safety analysis, startup, operation support, inspection, fitness for services and integrity analysis, retrofits and repairs. George has held engineering and management positions at Westinghouse and Washington Group International, where he has performed work at power and process plants, and consulted for the Department of Energy (DOE), the Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI).

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