Document: NUREG-0800
Document ID: dcd79479-0633-4d67-ba17-e369513aa55d
Document Type: srp
Title: SPECTRUM4 OF ROD EJECTION ACCIDENTS (PWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350410.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.8
CFR Part: 
CFR Title: 

Content:
and consequences acceptable. Since the calculations resulted in peak fuel enthalpies less than 280 cal/gm, prompt fuel rupture with consequent rapid heat transfer to the coolant from finely dispersed molten U02 was assumed not to occur. The pressure surge was, therefore, calculated on the basis of conventional heat transfer from the fuel and resulted in a pres- sure increase below "Service Limit C" (as defined in Section III, "Nuclear Power Plant Components," of the.ASME Boiler and Pressure Vessel Code) for the maximum control rod worths assumed. The staff believes that the calculations contain sufficient conservatism, both in the initial assumptions and in the analytical models, to ensure that primary system integrity will be maintained. V. IMPLEMENTATION The following section is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP Section. Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations. Implementation schedules for conformance to parts of the method described herein are contained in the referenced regulatory guide. VI. REFERENCES 1. 10 CFR Part 50, Appendix A, General Design Criterion 28, "Reactivity Limits." 2. Regulatory Guide 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors." 3. ASME Boiler and Pressure Vessel Code, Section III, "Nuclear Power Plant Components," American Society of Mechanical Engineers. 4. L. S. Tong, "Prediction of Departure from Nucleate Boiling for an Axially Non-Uniform Heat Flux Distribution," Jour. Nuclear Energy, Vol. 21, 241-248 (1967). 5. J. S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, and L. J. Stanek, "Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water," in "Two-Phase