Document: NRC Regulatory Guide
Document ID: 61b22006-634a-415a-a204-22e515d96707
Document Type: regulatory_guide
Title: Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents + HISTORY - HISTORY 11/2016 – DG-1327 , Proposed Revision 0
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML1612/ML16124A200.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.236
CFR Part: 
CFR Title: 

Content:
he conservatism of the models and codes should be evaluated both by comparison with experiment and with DG 1327, Page 8 more sophisticated spatial kinetics codes. In particular, the importance of two- or three- dimensional flux characteristics and changes in flux shapes should be investigated, and the conservatism of the flux shapes used for reactivity input and feedback, peak energy deposition, total energy, and gross heat transfer to the coolant should be evaluated. Also, sensitivity studies on variations of the Doppler effect, power distribution, fuel element heat transfer parameters, and other relevant parameters should be included. 2.1.2 The computer code used for calculating the transient should be a coupled thermal, hydrodynamic, and nuclear model with the following capabilities: (a) incorporation of all major reactivity feedback mechanisms, (b) at least six delayed neutron groups, (c) both axial and radial segmentation of the fuel element, (d) coolant flow provision, and (e) control rod scram initiation. 2.1.3 Calculations should be based upon design-specific information accounting for manufacturing tolerances. 2.1.4 Burnup-related effects on reactor kinetics (e.g., βeff, l*, rod worth, Doppler effect) and fuel performance (e.g., pellet radial power distribution, fuel thermal conductivity, fuel-clad gap conductivity, fuel melting temperature) should be accounted for in fuel enthalpy calculations. 2.2 Initial conditions 2.2.1 Accident analyses should be performed at beginning of cycle (BOC) and intermediate burnup intervals up to end of cycle (EOC). 2.2.2 Accident analyses at cold zero power (CZP) and hot zero power (HZP) conditions should encompass both (1) BOC following core reload and (2) re-start following recent power operation. 2.2.3 Accident analyses should be performed at intermediate power levels up to hot full power (HFP) conditions. These calculations should confirm power-dependent core operating limits (e.g., control rod insertion limits, rod