The specification of damping for nuclear piping systems subject to seismic-induced motions has been the subject of many studies and much controversy. Damping estimation based on test data can be influenced by numerous factors, consequently leading to considerable scatter in damping estimates in the literature. At present, nuclear industry recommendations and nuclear regulatory guidance are not consistent on the treatment of damping for analysis of nuclear piping systems. Therefore, there is still a need to develop a more complete and consistent technical basis for specification of appropriate damping values for use in design and analysis. This paper summarizes the results of recent damping studies conducted at Brookhaven National Laboratory.

ABSTRACT Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 and US NRC Standard Review Plan, Section 3.8) ; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 and 10 CFR 50); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 as well as SECY 90-016, SECY 93-087, and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.

The 100-40-40 rule is often used with the response spectrum analysis method to determine the maximum seismic responses from structural responses resulting from the three spatial earthquake components. This rule has been referenced in several recent Design Certification applications of nuclear power plants, and appears to be gaining in popularity. However, this rule is described differently in ASCE 4-98 and Regulatory Guide 1.92, consequently causing confusion on correct implementation of this rule in practice. The square root of the sum of the squares method is another acceptable spatial combination method and was used to justify the adequacy of the 100-40-40 rule during the development of the Regulatory Guide 1.92. The 100-40-40 rule, when applied correctly, is almost always conservative compared to the SRSS method, and is only slightly unconservative in rare cases. The purpose of this paper is to describe in detail the proper application of the 100-40-40 rule, as prescribed in ASCE 4-98 and in Regulatory Guide 1.92, and to clarify the confusion caused by the two different formats of this rule.