Document: NUREG-0800
Document ID: 0b17303b-e5cc-4091-b22c-056b0c78eb34
Document Type: srp
Title: – 15.3.4
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070707.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.3.3
CFR Part: 
CFR Title: 

Content:
100 guidelines. 4. The integrity of the reactor coolant pumps should be maintained such that loss of ac power and containment isolation will not result in pump seal damage. 5. The auxiliary feedwater system must be safety grade and, when required, automatically initiated. 6. Tripping of the reactor coolant pumps should be consistent with the resolution to Action Item II.K.3.5 of NUREG-0718 and NUREG-0737.31 76. A rotor seizure or shaft break in a reactor coolant pump should not, by itself, generate a 32 more serious condition or result in a loss of function of the reactor coolant system or containment barriers. 87. Only safety-grade equipment should be used to mitigate the consequences of the event. Safety functions should be accomplished assuming the worst single failure of a safety system active component (see Refs. 5 and 6). For new applications, loss of offsite power (LOOP) should not be considered a single failure; reactor coolant pump rotor seizures and shaft breaks should be analyzed with a LOOP (see item 9, below) in combination with a single active failure. (This position is based upon interpretation of GDC 17, as 15.3.3-5 DRAFT Rev. 3 - April 1996 documented in the Final Safety Evaluation Report for the ABB-CE System 80+ design certification.)33 98. The ability to achieve and maintain long-term core cooling coolability of the core34 should be verified. 109. This event should be analyzed assuming turbine trip and coincident loss of offsite power and coastdown of undamaged pumps. The applicant's analysis should be performed using an acceptable analytical model. The equations, sensitivity studies, and models described in References 8 through 12 are acceptable. The NRC staff found References 13 and 14 to be acceptable transient analysis computer codes for design analysis of the Advanced Boiling Water Reactor (ABWR). References 15 through 35 19 were found to be acceptable computer codes for transient analyses (i.e., except for loss-of- coolant accidents, or