Document: NUREG-0800
Document ID: 7b7303eb-a3a7-433b-8301-fcaba03194ea
Document Type: srp
Title: - 15.1.4
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070676.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.1.1
CFR Part: 
CFR Title: 

Content:
perational occurrences that have the potential to exceed allowable thermal design criteria for fuel cladding integrity. These four anticipated operational occurrences involve the transient increase in heat removal by the secondary system, which in turn causes reactor power to increase in response to the resultant lowering of the temperature of the reactor coolant. Regulatory Guide 1.53 provides guidance with respect to the application of the single failure criterion to the design and analysis of nuclear power plant protection systems. Regulatory Guide 1.105 provides guidance for ensuring that instrument setpoints are initially within and remain within the technical specification limits. Meeting the requirements of GDC 10 provides assurance that specified acceptable fuel design limits are not exceeded for the four anticipated operational occurrences evaluated in this SRP section involving excessive heat removal by the secondary system.36 (b) Compliance with GDC 15 requires that the reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient margin to ensure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences. GDC 15 is applicable to this section because the four overcooling events cause the reactor coolant system pressure to change in response to the drop in reactor coolant temperature. Although most of these events cause the reactor coolant pressure to decrease, some cause reactor coolant pressure to increase, depending on the worst single failure assumed. For example, for the ABWR the most severe initiating event in this group is a feedwater controller failure during maximum demand (runout of two feedwater pumps). This results in an increase in reactor pressure, but the increase is well within the ASME Code limit. Therefore, for the four overcooling transients of SRP Section 15.1.1, the reactor