Document: NRC Regulatory Guide
Document ID: 74c49394-8dbf-46e7-b62a-b85de93b47d8
Document Type: regulatory_guide
Title: Initial Test Programs for Water-Cooled Nuclear Power Plants + HISTORY - HISTORY 11/2012 – DG-1259 , Proposed Revision 4 11/2006 – DG-1166 , Proposed Revision 3 (Rev. 4)
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML1229/ML12298A071.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.68
CFR Part: 
CFR Title: 

Content:
hroughout core life, with the greatest worth control rod stuck out of the core. c. Verify adequate overlap of source- and intermediate range neutron instrumentation. d. Verify that proper operation of associated protective functions and alarms provides for plant protection in the low power range (if not previously performed). e. Verify flux distribution for comparison with distribution assumptions or predictions to check for potential errors in the loading or enrichment of fuel or lumped poison elements, as well as mis-positioned or uncoupled control rods. The measurements may be performed at a higher power level, depending on the sensitivity of in-core flux instrumentation using neutron and gamma radiation monitor surveys. Appendix A to DG-1259, Page A-23 f. Verify proper response of process and effluent radiation monitors. To the extent practical, responses by installed process and effluent radiation monitors should be verified by laboratory analyses of samples from the process and/or effluent systems. g. Verify chemical and radiochemistry tests and measurements to demonstrate the design capability of chemical control systems and installed analysis and alarm systems to maintain water quality within limits in the reactor coolant and secondary coolant systems. h. Verify neutron and gamma radiation surveys are performed. i. Demonstrate operability of control rod withdrawal and insertion sequencers and control rod withdrawal inhibit or block functions over the reactor power level range during which such features must be operable (BWR). j. Demonstrate the capability of the primary containment ventilation system to maintain the containment environment and important components in the containment within design limits with the reactor coolant system at rated temperature and with the minimum availability of ventilation system components for which the system is designed to operate. k. Demonstrate the operability of steam-driven engineered safety features, plant auxiliaries,