Plant states considered for the design
The category of plant states considered for the design of SSCs (also called plant design envelope in TECDOC-1791) in NS-R-1 includes the following:
• Normal Operation.
This operational state includes approach to criticality, start-up, power operation (steady operation and load following), normal shutdown, fuel handling, maintenance and testing.
The plant state of Normal Operation is usually subdivided in numbered 'modes' as follows:
1. power operation;
2. start-up;
3. hot standby, i.e. subcritical but approaching operating temperature;
4. hot shutdown, i.e. subcritical and temperature decreasing to below 100 ºC;
5. cold shutdown.
In mode 5, RCS as well as the containment may be open.
A more detail description of the various modes of operation for PWR and BWR is in EPRI TBR, section 2.2.2.1. Read more →
• Anticipated Operational Occurrences, which represent operational processes deviating from normal operation and which are expected to occur at least once during the operating lifetime of a facility but which, in view of appropriate design provisions, do not cause any significant damage to items important to safety or lead to accident conditions.
• Design Basis Accidents, which are not expected to occur during the plant life but are postulated to occur, so that calculated releases are within acceptable limits, using established design criteria and conservative methodology. Design Basis Accidents represent conservative set of bounding accidents determined with pessimistic assumptions on the evolution of the sequence and for which SSCs important to safety are designed;
An example of a DBA is a large break of a main reactor coolant pipe, leading to a large Loss of Coolant Accident (Large Break LOCA). It should be noted that DBAs may lead to some fuel damage (e.g., ballooning and rupture of fuel cladding), but core coolability is not lost.
• Accidents more severe than DBAs are called Beyond Design Basis Accidents (BDBAs). These include both accidents without significant core degradation and so-called Severe Accidents (SA), which involve significant core degradation. Sequences that may lead to SAs are identified and a few sequences are considered in the design, using deterministic and probabilistic methods and engineering judgement, for the definition of design measures and procedures to deal with the selected sequences (in particular for accident management).
For each of the plant states, acceptance criteria are defined, from which design criteria for the mitigating systems are derived. This subject is treated in Subchapter 3.1.3.
This results in the design of Structures, Systems and Components (SSCs), which fulfil the fundamental safety functions and the other safety functions, derived from these, as described in Subchapter 2.3.
Now, SSR-2/1 (Rev. 1) Requirement 13 also states that plant states shall be identified and shall be grouped into a limited number of categories according to their frequency of occurrence at the nuclear power plant. Similarly to NS-R-1, Req. 13 also states that criteria shall be assigned to each plant state, such that frequently occurring plant states shall have no, or only minor, radiological consequences and plant states that could give rise to serious consequences shall have a very low frequency of occurrence.
Plant states identified in this requirement typically cover:
• Normal operation;
• Anticipated operational occurrences, which are expected to occur over the operating lifetime of the plant;
• Design basis accidents;
• Design extension conditions, including accidents with core melting.
Hence the fundamental difference with NS-R-1 is that SSR-2/1 (Rev. 1) states the consideration of
Design Extension Conditions (DEC) – including accidents with core melting – instead of Beyond Design Basis Accidents.
The principle of Design Extension Conditions, and their consideration in the design process and related safety assessment, is further developed in a variety of requirements, including Req. 7, 20 and 33. These requirements clarify that the notion of DEC goes far beyond a simple matter of terminology (where Beyond Design Basis Accidents in NS-R-1 would be simply renamed
Design Extension Conditions in SSR-2/1 (Rev. 1), to imply a
far-reaching extension of the plant states considered
for the design, to fully include DEC within the design
basis (as opposed to BDBA in NS-R-1).
In fact, Req. 20 of SSR-2/1 (Rev. 1) states that “a set of design extension conditions shall be derived on the basis of engineering judgement, deterministic assessments and probabilistic assessments for the purpose of further improving the safety of the nuclear power plant by enhancing the plant’s capabilities to withstand, without unacceptable radiological consequences, accidents that are either more severe than design basis accidents or that involve additional failures. These design extension conditions shall be used to identify the additional accident scenarios to be addressed in the design and to plan practicable provisions for the prevention of such accidents or mitigation of their consequences.
The design extension conditions shall be used to define the design specifications for safety features and for the design of all other items important to safety that are necessary for preventing such conditions from arising, or, if they do arise, for controlling them and mitigating their consequences.
The analysis undertaken shall include identification of the features that are designed for use in, or that are capable of preventing or mitigating, events considered in the design extension conditions. These features:
a) Shall be independent, to the extent practicable, of those used in more frequent accidents;
b) Shall be capable of performing in the environmental conditions pertaining to these design extension conditions, including design extension conditions in severe accidents, where appropriate;
c) Shall have reliability commensurate with the function that they are required to fulfil.
In particular, the containment and its safety features shall be able to withstand extreme scenarios that include, among other things, melting of the reactor core. These scenarios shall be selected by using engineering judgement and input from probabilistic safety assessments”.
The broadening of the design plant envelope from NS-R-1 to SSR-2/1 (Rev. 1) is represented schematically in the Figure 1-3.

Figure 1-3: Plant design envelope in SSR-2/1 (Rev. 1), broadened with respect to NS-R-1, to include Design Extension Conditions.
Read more:
SSR-2/1 (Rev. 1),
TECDOC-1791
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