Document: NUREG-0800
Document ID: a2598df1-7ec6-43a6-9d2a-d8210d1f944f
Document Type: srp
Title: through 7.9. Additional information relevant to the review process can be found in the references in
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0525/ML052500461.pdf
Revision Date: 2023-06
Chapter: 7
Section ID: 7.1
CFR Part: 
CFR Title: 

Content:
nce, and on-line testing. The review process for these topics must recognize the special characteristics of digital systems. C. Review Process C.1. Summary The overall process for reviewing the unique aspects of digital I&C systems is outlined in Figure 7.0-A-1. Figure 7.0-A-2 shows the issue-resolution process applicable to each item in 7.0-A-1. The process shown in Figure 7.0-A-1 applies to any digital I&C system or function proposed in a license application or a license amendment application. The scope of the review process is the same for any I&C safety function; however, the effort required to implement the review will be considerably less for a system that implements only a few safety requirements than it will be for a complex system such as a complete, integrated, digital safety system design. While The Staff discussed the issues of classification and requirements grading in SECY-91-292, "Digital Computer Systems for 1 Advanced Light-Water Reactors," and noted that, "A graded set of requirements based on the importance to safety of the functions being performed with respect to reduction in the potential for radiation exposure could be adopted." IEEE Std 603 and IEEE Std 7- 4.3.2, "IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations," endorsed by Reg. Guide 1.153 and Reg. Guide 1.152, "Criteria for Digital Computers in Safety Systems of Nuclear Power Plants," do not provide for classification, although the foreword to IEEE Std 7-4.3.2 recommends the addition of grading to future versions of IEEE Std 603. Rev. 4 — June 1997 SRP 7.0-A-5 acceptance criteria remain the same, the Staff's review emphasis should be commensurate with the safety 1 significance of the given system or aspect of a system's design under review. Probabilistic risk assessments (PRAs), such as those conducted under the Individual Plant Evaluation program (see Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities") or