Document: NUREG-0800
Document ID: 061b3f99-1bfd-4e65-be97-7e9affac9aef
Document Type: srp
Title: UNCONTROLLED CONTROL ROD ASSEMBLY WITHDRAWAL AT POWER
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0636/ML063600414.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.2
CFR Part: 
CFR Title: 

Content:
correctly calculated. The range of parameters to be considered includes: A. Initial power levels from low to full power. B. Reactivity insertion rates from very low to maximum possible for the control system, including allowance for uncertainties. C. Fuel and moderator feedback reactivity coefficients covering the range expected throughout the cycle, including allowance for uncertainties. D. Power peaking factors at design limits for the initial power level conditions. 4. For both types of reactors, the reviewer determines whether the applicant's analytical methods and models are acceptable, including steady-state, AOO, system response, and fuel response models. This may be done by using one or more of the following procedures: A. Determine whether the method has been reviewed and approved previously by considering past safety evaluation reports (SERs) and reports prepared in response to technical assistance requests. B. Perform an independent review of the method (usually described in a separate licensing topical report and frequently completed, on a generic basis, outside the scope of the review for a particular facility). C. Perform auditing-type calculations using methods available to the staff. D. Request additional bounding calculations from the applicant to confirm the validity of those portions of the applicant's analytical methods that are not fully reviewed or approved. 15.4.2-7 Revision 3 - March 2007 5. For new application reviews, the analysis must consider a loss of offsite power in conjunction with the limiting single active failure when assessing the consequences of the anticipated operational occurrence. (This position is based upon interpretation of GDC 17, as documented in the Final Safety Evaluation Report for the ABB-CE System 80+ design certification, NUREG-1462, Volume 2, August 1994). 6. The results of the analysis should be presented and should include maximum power levels reached for the reactor and the peak fuel rod, scram or rod block