Document: NUREG-0800
Document ID: 7abc5861-4ac1-4f27-ab53-f9997c382301
Document Type: srp
Title: SPECTRUM OF ROD EJECTION ACCIDENTS (PWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0705/ML070550014.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.8
CFR Part: 
CFR Title: 

Content:
is radiological consequence analyses associated with design basis accidents are reviewed under SRP Section 15.0.3. 3. Reactivity coefficients and control rod worths are reviewed under SRP Section 4.3. 4. Relevant thermal-hydraulic analyses are reviewed under SRP Section 4.4. 5. The applicant's determination of the reactor trip delay time (i.e., the time elapsed between when the sensed parameter reaches the level for which protective action is required and the onset of negative reactivity insertion) is reviewed under SRP Sections 7.2 and 7.3. The specific acceptance criteria and review procedures are contained in the referenced SRP sections. II. ACCEPTANCE CRITERIA Requirements Acceptance criteria are based on meeting the relevant requirements of the following Commission regulations: 1. General Design Criterion (GDC) 13, as to the availability of instrumentation to monitor variables and systems over their anticipated ranges to assure adequate safety, and of appropriate controls to maintain these variables and systems within prescribed operating ranges. 2. Acceptance criteria are based on meeting GDC 28 requirements as to the effects of postulated reactivity accidents that result in neither damage to the reactor coolant pressure boundary greater than limited local yielding nor sufficient damage to impair significantly core cooling capacity. Regulatory positions and specific guidelines necessary to meet the relevant requirements of GDC 28 are in Regulatory Guide 1.77 and SRP Section 4.2. 15.4.8-3 Revision 3 - March 2007 The maximum reactor pressure during any portion of the assumed excursion should be less than the value that result in stresses that exceed the "Service Limit C" as defined in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. 3. 10 CFR 100.11 and 10 CFR 50.67 establish radiation dose limits for individuals at the boundary of the exclusion area and at the outer boundary of the low population zone. The fission product