Document: NUREG-0800
Document ID: 6997f9d9-dedc-49d7-ac15-7615c1bd9713
Document Type: srp
Title: SPECTRUM OF ROD DROP ACCIDENTS (BWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0705/ML070550015.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.9
CFR Part: 
CFR Title: 

Content:
e pressure surge results in a pressure increase below "Service Limit C" (as defined in Section III of the ASME 15.4.9-6 Revision 3 - March 2007 Boiler and Pressure Vessel Code) for the maximum control rod worths assumed. The staff believes that the calculations are sufficiently conservative in both the initial assumptions and analytical models to maintain primary system integrity. For DC and COL reviews, the findings will also summarize (to the extent that the review is not discussed in other SER sections) the staff's evaluation of the ITAAC, including design acceptance criteria, as applicable, and interface requirements and combined license action items relevant to this SRP section. V. IMPLEMENTATION The staff will use this SRP section in performing safety evaluations of DC applications and license applications submitted by applicants pursuant to 10 CFR Part 50 or 10 CFR Part 52. Except when the applicant proposes an acceptable alternative method for complying with specified portions of the Commission’s regulations, the staff will use the method described herein to evaluate conformance with Commission regulations. The provisions of this SRP section apply to reviews of applications submitted six months or more after the date of issuance of this SRP section, unless superseded by a later revision. VI. REFERENCES 1. 10 CFR 50, Appendix A, GDC 13, “Instrumentation and Control.” 2. 10 CFR 50, Appendix A, “General Design Criteria for Nuclear Power Plants,” GDC 28, "Reactivity Limits." 3. NUREG-1503, ABWR Final Safety Evaluation Report, Section 15.4.1, "Control Rod Drop Accidents," July 1994. 4. ASME Boiler and Pressure Vessel Code, Section III, "Nuclear Power Plant Components," American Society of Mechanical Engineers. 5. "Rod Drop Accident Analysis for Large Boiling Water Reactors," NED0-10527, General Electric Company, March 1972; Supplement 1 to NED0-10527, July 1972; and Supplement 2 to NED0-10527, January 1973. PAPERWORK REDUCTION ACT STATEMENT The