Document: NUREG-0800
Document ID: 5c84a1c2-9405-4849-8c5f-415a0bb66a22
Document Type: srp
Title: in connection with reactor internals.
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070337.pdf
Revision Date: 2023-06
Chapter: 3
Section ID: 3.9.5
CFR Part: 
CFR Title: 

Content:
, December 20, 1988.108 14. NUREG/CR-5416, "Technical Evaluation of Generic Issue 113: Dynamic Qualification and Testing of Large Bore Hydraulic Snubbers"; Nitzel, M. E.; Ware, A. G. EG&G Idaho, Inc.; Page, J. D. NRC; September 1992 (EGG-2571).109 3.9.3-19 DRAFT Rev. 2 - April 1996 APPENDIX A STANDARD REVIEW PLAN SECTION 3.9.3 STRESS LIMITS FOR ASME CLASS 1, 2, AND 3 COMPONENTS AND COMPONENT SUPPORTS, OF SAFETY-RELATED SYSTEMS AND CLASS CS CORE SUPPORT 110 111 STRUCTURES UNDER SPECIFIED SERVICE LOADING COMBINATIONS A. INTRODUCTION Nuclear power plant components and supports are subjected to combinations of loadings 112 derived from plant and system operating conditions, natural phenomena, postulated plant events, and site-related hazards. Section III, Division 1 of the ASME Code (hereafter referred to as the Code) provides specific sets of design and service stress limits that apply to the pressure retaining or structural integrity of components and supports when subjected to these loadings.113 Conditions also warranting consideration include thermally stratified flow, thermal striping, and/or thermal cyclic effects and the resulting spatial or temporal stresses on piping and components. Such conditions, where not identified and accounted for in stress analysis and fatigue evaluations of affected piping, can result in unacceptable stresses, pipe movements, deformations, and/or fatigue failures. These phenomena have typically been observed in feedwater piping, at feedwater nozzles, in PWR pressurizer surge lines, in piping between PWR pressurizers and associated relief valves, and at locations where piping normally containing relatively cool fluid is connected to the reactor coolant system via valves subject to intermittent leakage such as at residual heat removal and emergency core cooling system connections to the reactor coolant system (References 12 and 13).114 The design and service stress limits specified by the Code do not assureensure, in themselves, the