Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
49 Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
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- Ris (Zotero)
- Reference Manager
In probabilistic safety assessment (PSA) of nuclear power plants, operator actions to prevent core damage are evaluated using well established human reliability analysis (HRA) methods. These operator actions analyzed are performed within a clear boundary: There is a well defined operator crew that uses clearly written and obliging procedures, the so called Emergency Operating Procedures (EOPs). Under these circumstances, the determination of human error rates using well established HRA methods is relatively straight forward. Such type of operator actions are modeled in the Level 1 part of probabilistic safety assessment studies.
There is a different situation for actions to mitigate severe accidents. Severe Accident Management Guidance (SAMG) was developed by reactor vendors in the 1990s. SAMGs give guidance to the plant staff to maintain containment integrity and to minimize radiological releases in case a potential core damage accident has occurred. Therefore, any SAMG actions are modeled in the containment event tree of the Level 2 part of a probabilistic safety study.
There are two main differences of SAMGs compared to EOPs: SAMGs represent guidance and evaluations, but they do not represent procedures. And SAMGs are used by the plant emergency staff including the whole emergency organization and not by the operator crew only. Therefore, the boundary conditions of SAMG actions are quite different than those of classical HRA methods for EOP operator actions.
SAMG were implemented at the Beznau Nuclear Power Plant in 2000. The Beznau plant is located in Northern part of Switzerland and consists of two Westinghouse pressurized water reactors. Several accident management measures were implemented at the Beznau plant. Examples are the construction of a containment filtered vent system and the installation of several fire water connections for injection using mobile pumps.
To support the alignment of mobile equipment for accident management purposes, an additional set of procedures was developed at Beznau, the so-called Accident Management Procedures (AMP). They advise the relevant emergency crews such as operations, fire water and others to align mobile equipment as required by the plant emergency staff. These Accident Management Procedures require coordination between at least two different emergency crews. Also for these type of AMP operator actions, the boundary conditions such as coordination of different emergency crews are different that for classical EOP actions.
As a result, when actions according to the Severe Accident Management Guidance or according to the Accident Management Procedures are modeled in the PSA study, methods different the classical human reliability analysis methods have to be used.
In this paper, a simplified method is shown to evaluate and implement human error rates of any type of accident management actions, which are beyond of the scope of the emergency operating procedures, in the PSA study.
The methodology distinguishes between errors of the emergency staff when working with the SAMGs and execution errors of the operations team during execution of actions. Examples of more complex operator actions are the alignment of mobile equipment such as fire water pumps.
The methodology performs a very rough calculation of human error rates using only three different human failure rates. But the methodology considers dependencies between different human failures and also assumes a common human failure rate for the entrance into the SAMG package by of the emergency staff.
The overall approach allows calculation of the benefit of an implemented accident management program on nuclear power plant safety.