Document: NUREG-0800
Document ID: 94ab38ac-ddfc-4ba1-ae38-bc25ffa6e976
Document Type: srp
Title: – 15.4.5
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070716.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.4
CFR Part: 
CFR Title: 

Content:
dents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor temperature and pressure, and cold water addition. GDC 28 is applicable to this section because the reviewer evaluates the consequences of the events associated with startup of an inactive loop or a recirculation loop at an incorrect temperature and with a flow controller malfunction causing an increase in BWR core flow rate. This section, SRP Sections 4.2 through 4.4 and 7.2 through 7.5, and Regulatory Guides 1.53 and 1.105 provide guidance for ensuring that the reactor coolant system and associated auxiliary, control, and protection systems are designed with appropriate margin to ensure that the reactor coolant pressure boundary will not be breached. Meeting the requirements of GDC 28 provides assurance that the reactor coolant pressure boundary will not be breached during any condition of normal operation, including the effects of AOOs.31 III. REVIEW PROCEDURES The procedures below are used during both the construction permit (CP), and operating license (OL), and combined license (COL) reviews. During the CP review, the values of system parameters and setpoints used in the analysis will be preliminary in nature and subject to change. At the OL or COL review stage, final values should be used in the analysis, and the reviewer 32 should compare these to the limiting safety system settings included in the proposed technical specifications. DRAFT Rev. 2 - April 1996 15.4.4-8 The description of the core flow increase transients(AOOs) presented in the SAR is reviewed by RSBSRXB regarding the occurrences leading to the initiating event. The sequence of events from initiation until a stabilized condition is reached is reviewed to ascertain: 1. The extent to which normally operating plant instrumentation and controls are assumed to function. 2. The extent to which plant and reactor protection systems are required to function. 3. The