Document: NUREG-0800
Document ID: 3b714304-f91e-4999-917a-7d27866735e3
Document Type: srp
Title: STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT (PWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350117.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.1.5
CFR Part: 
CFR Title: 

Content:
ly detailed to simulate the reactor coolant (primary), steam generator (secondary), and auxiliary systems. The reviewer evaluates the following functional requirements: (1) Reactor trip signal: credit taken for any reactor trip signal is reviewed by ICSB to confirm that, under accident conditions, the instrumentation and control systems are capable of the assumed response. (2) Emergency core cooling system (ECCS): credit taken for actuation of the ECCS is reviewed by ICSB to verify the ability of the instrumentation and control systems to respond as assumed. (3) Auxiliary feedwater system: the availability of the auxiliary feedwater system to supply adequate auxiliary feedwater flow to the intact steam generators during the accident and the subsequent shut- down condition is evaluated. This is done by ASB as to availability of the system and by RSB as to capability to effect an orderly shutdown. Since auxiliary feedwater system designs are diverse and may require both automatic and manual actuation, preoperational tests should be specified to identify any necessary operator actions and to establish times required for their completion. d. The variations with time during the transient of the neutron power, heat fluxes (average and maximum), total core reactivity, reactor coolant system pressure, minimum DNBR; coolant conditions (inlet temperature, core average temperature and average exit and hot channel exit temperatures, fuel rod conditions (maximum fuel center- line temperature, maximum clad temperature, or maximum fuel enthalpy), steam generator pressure, containment pressure, relief and/or safety valve flow rates, discharge flow rate, steam line and feedwater flow rates and pressurizer and steam generator water levels are reviewed. The values of the more important of these parameters for the steam line break accident (as listed in subsection I) are compared with those predicted for other similar plants to see that they are within the range expected. 2. To the