Document: NUREG-0800
Document ID: e6b7e6d3-a99d-424d-bfa0-491bc48f46cf
Document Type: srp
Title: ISOLATION CONDENSER SYSTEM (BWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0708/ML070810517.pdf
Revision Date: 2023-06
Chapter: 5
Section ID: 5.4.13
CFR Part: 
CFR Title: 

Content:
le by compliance with the applicable provisions of the ASME Code and by compliance with the positions of Regulatory Guide 1.44, “Control of the Use of Sensitized Stainless Steel.” Regulatory Guide 1.44 contains staff positions related to unstabilized austenitic stainless steel of the AISI Type 3XX series used for components of the RCPB. Positions related to BWR piping materials, including verification of nonsensitization of the material by an approved test, are described in Attachment A to Generic Letter 88-01. The technical bases for the positions provided in Generic Letter 88-01 and similar recommendations related to minimizing stress corrosion cracking in susceptible piping of BWRs are detailed in NUREG-0313, Revision 2. Upon resolution of GSI-191, the review should include consideration of the resolution of this issue. 3. Pursuant to GDC 5, SSCs that are important to safety should not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, the event of an accident in one 5.4.13-7 March 2007 unit, an orderly shutdown and cooldown of the remaining units. With respect to GDC 5, the application should demonstrate that the ICS design’s ability to accomplish these safety-related functions is not compromised for each unit regardless of equipment failures or other events that may occur in another unit. 4. With respect to GDC 17, the application should demonstrate conformance with the guidelines in RG 1.93 with respect to providing onsite and offsite electric power systems to permit functioning of SSCs important to safety to ensure their safety function assuming either power system is not functioning. The application should demonstrate sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences