Document: NUREG-0800
Document ID: d5452e7b-1e61-498b-9e3d-71073b3328ef
Document Type: srp
Title: LEAK-BEFORE-BREAK EVALUATION PROCEDURES
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0302/ML030280295.pdf
Revision Date: 2023-06
Chapter: 3
Section ID: 3.6.3
CFR Part: 
CFR Title: 

Content:
ction (for example, when snubbers are reduced in number or capacity in older operating plants; on the other hand, changing high strength fastener material would not require the use of current codes or NRC criteria). In heavy component support redesign, the already existing SSE may be used, and improved functional reliability must be 3.6.3-10 demonstrated for any changes implemented. Structural capacity associated with the original steel and concrete, including struts, columns, pedestals, hangers, trusses and skirts cannot be diminished in the support system of operating plants or plants under construction. Redesign will be limited to replacing high strength fastener material and reducing the number and capacity of snubbers. Applicants and licensees undertaking heavy component support redesign, with dynamic effects of pipe rupture eliminated, should use in dependent design and fabrication verification procedures to minimize the potential for design and construction errors. Displacements and rotations resulting from potential failure of redesigned lateral (horizontal) supports should not lead to the rupture of piping connected to the reactor coolant loop heavy components. VI. REFERENCES 1. 10 CFR Part 50, Appendix A, General Design Criterion 4, "Environmental and Dynamic Effects Design Bases". 2. NUREG-1061, Volume 3, "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks", November 1984 . 3. Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems". 4. EPRI Report NT-4690-SR, "Evaluation of Flaws in Austenitic Steel Piping", April, 1986. 3.6.3-11