Document: NUREG-0800
Document ID: a7fafe76-5570-446b-96dd-3e53ec9bccb3
Document Type: srp
Title: NUREG-0800
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052340663.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4
CFR Part: 
CFR Title: 

Content:
um worths of individual rods or banks as a function of position for power and cycle life conditions appropriate to rod withdrawal transients and rod ejection or drop accidents. Descriptions and curves of maximum rates of reac- tivity increase associated with rod withdrawals, experimental confirm- ation of rod worths or other factors justifying the reactivity increase rates used in control rod accident analyses, and equipment, adminis- trative procedures, and alarms which may be employed to, restrict potential rod worths should be included. d. Descriptions and graphs of scram reactivity as a function of time after scram initiation and other pertinent parameters, including methods for calculating the scram reactivity. 6. The area of criticality of fuel assemblies. Discussions and tables giving values of K for single assemblies and groups of adjacent fuel assemblies up to the naer required for criticality, assuming the assemblies are dry and also immersed in water, are reviewed. 7. The areas concerning analytical methods. These are: a. Descriptions of the analytical methods used in the nuclear design, including those for predicting criticality, reactivity coefficients, burnup, and stability. 4.3-3 Rev. 2 - July 1981 b. The data base used for neutron cross-sections and other nuclear parameters. c. Verification of the analytical methods by comparison with measured data. 8. The areas concerning pressure vessel irradiation. These are: a. Neutron flux spectrum above 1 MeV in the core, at the core boundaries, and at the inside pressure vessel wall. b. Assumptions used in the calculations; these include the power level, the use factor, the type of fuel cycle considered, and the design life of the vessel. c. Computer codes used in the analysis. d. The data base for fast neutron cross sections. e. The geometric modeling of the reactor, support barrel, water annulus, and pressure vessel. f. Uncertainties in the calculation. 9. The adequacy of limits on power distribution during