CHAPTER 3: INTRODUCTION TO ACCIDENT MANAGEMENT

Classification of Structures, Systems & Components

Introduction

The need to classify equipment in a nuclear power plant according to its importance to safety has been recognized since the early days of reactor design and operation.
The method for classifying the safety significance of items important to safety shall be based primarly on deterministic methodologies complemented where appropriate, by probabilistic methods, taking into account factors such as:

• The safety functions to be performed by the item.
• The consequences of failure to perform the safety function.
• The frequency with which the item will be called upon to perform a safety function.
• The time following a postulated initiating event at which, or the period of which, the item will be called upon to perform a safety function.

Further information can be found in SSR 2/1, Rev. 1, Requirement 22 Read more →

SSG-30
Figure 1-4: Flow chart indicating the classification process, SSG-30.

Classification of Systems, Structures and Components

Once the categorization of functions is completed, and SSC are assigned to one or more safety functions, these SSC should be classified for safety. A widely used example of assignment of classes is as follows, where classifications should be established and should be verified for adequacy and consistency:

• Classification of systems on the basis of the importance of the affected safety function;
• Classification for pressure components, on the basis of the severity of the consequences of their failure, mechanical complexity and pressure rating;
• Classification for resistance to earthquakes, on the basis of the need for the structure or component considered to retain its integrity and to perform its function during and after an earthquake, taking into account aftershocks and their consequential incremental damage;
• Classification of electrical, instrumentation and control systems on the basis of their safety or safety support functions, which may be different from the classification of other plant systems owing to the existence of field specific, widely used classification schemes;
• Classification for quality assurance provisions

Further information on classification methodology is available in SSG-30, Read more →

How the various classifications of the SSC can be interrelated, can be seen in the following Table 1-3.

Safety  Pressure integrity 1  Electrical 2  Seismic  Environmental  QA 
Safety 3 
High class  High class  High category  Harsh, mild  High class 
Supplemented grade 4 
Intermediate class  Low class or no requirements  Low category or no requirements  Various, or no requirements  Low class 
Non-safety 
Non-nuclear  No requirements  No requirements  No requirements  No requirements 


1 Widely used is classification according to the ASME III code, the RCCM code, other similar codes for pressure integrity (ASME = American Society of Mechanical Engineers; RCCM = French equivalent)
2 Examples in IEEE standard 323 (IEEE = Institute of Electric and Electronic Engineers)
3 In this example (simple sketch only - details are in the ANS/ANSI Standard 58.14-2011) there is only one safety class; often, more safety classes are defined, e.g. safety classes 1-3, as is the case in the still widely used ANS 51.1 and ANS 52.1, in Russian safety guides and in the IAEA SSG-30. The requirements may then be further subdivided (ANS = American Nuclear Society, ANSI = American National Standards Institute).
4Supplemented grade is not ´safety´, but has a significant licensing requirement or commitment; could also be used for equipment needed to mitigate DECs (is then ´safety` in SSG-30).

Table. 1-3 - Example of scheme of interrelation between safety classes for various parts of equipment and the associated requirements

The IAEA has published the Special Safety Guide on Safety System Classification SSG-30 in 2012, i.e long after most NPP designs have been completed and the plants been built. Therefore, NPPs operating today (2020) meet national or international safety standards for classification, rather than SSG-30. In addition, SSG-30 is intended to be applied to new plants. See para. 1.9 of SSG-30. Read more →

For example, many plants have followed the classification according to ANS 51.1 (PWRs, 1983) and ANS 52.1 (BWRs, 1983). These standards have been superseded by ANS 58.14 (both PWRs and BWRs, 1993, revised 2012 to ANS 58.14-2011, reconfirmed 2017). The classification standards ANS 51.1, ANS 52.1 and ANS 58.14-2011 are.available from the American Nuclear Society, www.ans.org

These standards meet GSR Part 4 (Rev. 1), para 4.30.

New plants have usually dedicated hardware to mitigate DECs, including fuel damage accidents. This does not only affect the classification method, it also affects the SAMG. Therefore, in this SAMG-D, it is assumed that a national classification system is used. The method to develop SAMG as such, is not expected to change much with application of SSG-30.

A difference between most national standards and SSG-30 is that they do not classify systems designed to mitigate DECs (if any such system is present). In SSG-30, systems designed to mitigate DECs are Safety Class 2 or 3. An option in some present day standards is to classify those systems on the basis of a licensing commitment.

I.e., there is no regulatory or standards requirement, but a voluntary commitment by the licensee. The method is mentioned in e.g. ANS 58.14 (1993, 2012). A licensing commitment becomes a licensing requirement when it has been specified in the plant licensing documentation basis (LDB). Note that ANS 58.14 does not refer to application of the licensing commitment to systems designed to mitigate DECs.

See SSG-30, para 3.15, Safety Category 3, and para 3.16, including Table 1. Read more →