Document: NUREG-0800
Document ID: ef559b35-0671-46e4-b172-09155cca8b81
Document Type: srp
Title: SPECTRUM OF ROD DROP ACCIDENTS (BWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350427.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.9
CFR Part: 
CFR Title: 

Content:
the coolant from finely dispersed molted 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 pressure increase below "Service Limit C" (as defined in Section III 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. 15.4.9-3 Rev. 2 - July 1981 VI. REFERENCES 1. "Rod Drop Accident Analysis for Large General Electric Company, March 1972; 1972; and Supplement 2 to NEDO-10527, Boiling Water Reactors," NEDO-10527, Supplement 1 to NEDO-10527, July January 1973. 2. 10 CFR 50, Appendix A, General Design Criterion 28, "Reactivity Limits." 3. ASME Boiler and Pressure Vessel Code, Section III, "Nuclear Power Plant Components," American Society of Mechanical Engineers. I I 15.4.9-4 Rev. 2 - July 1981 NUREG-0800 (Formerly NUREG-75/087) Adso REG04 I\ .°, U.S. NUCLEAR REGULATORY COMMISSION <tt STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION