Accident Sequence Development

The reaction chain within a system can be broken (deliberate and planned intervention as an accident sequence develops to halt or reduce the progress of the event). 

Accident-sequence development was discussed in Section 2.6.4. The major differences in this step for external events as contrasted with traditional internal events are the addition of external event-caused failures to the fault trees and the increased likelihood of multiple failures of safety systems due to correlations between component responses and between component capacities. There are additional considerations when determining core damage frequencies associated with fires. These considerations include the availability and effectiveness of automatic and manual fire suppression, and the locations of vital equipment with respect to potential fires. Coincident failures of fire protection systems and other systems are also considered. Only a small fraction of the fires that could occur in a nuclear power plant would be expected to lead to core damage. 

A significant development of the study was the use of event trees to link the system fault trees to (lie accident initiators and the core damage states as described in Chapter 3. This was a response to the ditficulties encountered in performing the in-plant analysis by fault trees alone. Nathan Villalva and Winston Little proposed the application of decision trees, which was recognized by Saul Levine a.s providing the structure needed to link accident sequences to equipment failure. 

Since this text is not solely devoted to HRA, the above process is by necessity simplified. A more defined interaction is shown in Figure 4.5-2. Here HRA interacts with the development of the system models and feeds into not only the accident sequences but also the physical analysis of the inplant and explant accident progress,... 

Accident Sequence Modeling using a logic model for the facility was developed. The model included all initiators of potential accidents and the response of the installation to these initiators. Specific accident sequences are defined in event trees consisting of an initiating... 

In April 1982, a data workshop was held to evaluate, discuss, and critique data in order to establish a consensus generic data set for the USNRC-RES National Reliability Evaluation Program (NREP). The data set contains component failure rates and probability estimates for loss of coolant accidents, transients, loss of offsite power events, and human errors that could be applied consistently across the nuclear power industry as screening values for initial identification of dominant accident sequences in PRAs. This data set was used in the development of guidance documents for the performance of PRAs. 

Propagating events or propagating factors allow the accident sequence to continue, developing consequences that are subsequently more severe. Examples are failures of release-detection systems or the presence of ignition sources. 

Improved models of the accident sequence would be helpful in understanding the dynamics of accidents and would be a basis for data collection. No fuUy satisfactory model has yet been developed, but many are promising and useful [p. 85]. 

The development of a standard for performing LPS PSA should be considered. This recommendation is based on observations made during presentations of PSA results for plants of the same design and reactor size. For example, different initiating event lists, even to the point of defining different LOCA sizes, different definitions of POS and different dominant accident sequences were presented. The standard should be written to minimize variation among PSAs of reactors of the same design and size, but should ensure that plant-specific operational differences are reflected in the PSAs. 

Moreover, one of the most difficult aspects of the probabilistic analyses lies in the probabilistic treatment of human behaviour, that is of the operator actions which may have a decisive influence on the development of the accidental sequence under study. Usually, for the sake of conservatism, focus is placed on the probability of operator error (omission, commission and, more difficult to analyse, diagnosis errors). In the real world, however, the role of operators in an accident sequence is not limited to committing or not committing mistakes in the implementation of operational procedures. In fact, as many events indicate (the Browns Ferry 1975 accident is typical, see Chapter 20), the operators may react to an unexpected situation with creative and resolving interventions. For the present moment, however, except for specific cases, the possibility is taken into account only that the operator makes mistakes in the implementation of emergency procedures, even in the field of the management of severe accidents. 

Event tree analysis was used to analyze the accident sequence and to evaluate the accident frequency as suggested in (DOE 1997) and (Mahn et al. 1995). Figure 3E.3-1 presents the event tree using the two accident progression sequences developed previously. 

Symptom based procedures can be directly used as the best information source for building of accident scenarios models and accident sequences logic covering the phase of control room crew response to initiating event. According to current requirements, the symptom based procedures set developed for specific plant should cover response to all initiating events defined in plant PSA. 

Probabilistic codes which develop a logical model to identify the accident sequences which could occur following PIEs and estimate their frequencies. 

SECY-04-0210, Status of accident sequence precursor program and the development of standardized 2004 plant analysis risk models... 

ASEP Accident sequence evaluation program. It is almost THERE (discussed in later), but requires fewer resources. It is mainly developed in the United States for nuclear program. It can be used by non-HRA specialists. Both pre- and postaccident quantification of HFEs are done in ASEP. Usually, it gives conservative output. It is used for nuclear plants. 

Developing the actual (as opposed to what people believe occurred) sequence of events is crucial. Various safety analysis techniques discussed in Chapters 5 through 9 are excellent methods to use in attempting to reconstruct a timeUne of activities. These tools are also very useful for linking accident sequences to sequence causes. The analysis should clearly indicate primary, contributory, and other causal factors that led to the accident. 

Currently, the assessment of seismic capacity is being carried out to comply with the relevant IAEA recommendations. In addition, a probabilistic seismic hazard analysis is included in the Temelin PSA Project scope in order to address the contribution from earthquake induced accident sequences to the overall CDF of the plant. The seismic hazard curves for the Temelin site has been developed and seismic fragility analysis has been performed for the structures and components. Based on the preliminary results (annual frequency of O.lg PGA (SSE) earthquake is lE-6/year), it is expected that the contribution of seismic events and the consequential accident sequences to the overall CDF will be negligible (i.e. less then 1% of overall CDF). The independent review of this PSA task is to be carried out in the framework of the 2nd IAEA IPERS (Level 1 - external initiating events) in August/September 1995. 

Certain Beyond Design Basis accident sequences could lead to a core melt whilst this is extremely unlikely, it is not incredible. This possibility required some form of mitigation. The AP1 OOO s designers addressed this challenge by developing the capability for in-vessel retention. The alternative would have been to incorporate some form of core catcher outside the reactor vessel. A core catcher would have features that precluded re-criticality of the mixtures of core structural materials and building structures (known as corium), and cooled it to slow its reaction with materials around the reactor vessel. This could have been the design solution for APIOOO. 

The last topic was found strongly related to an option of NPP licensing with reduced or eliminated off-site emergency planning requirements. It was recommended to re-examine such option for the innovative SMRs developed currently by considering both the required institutional changes and the accident sequences that need to be eliminated to achieve this objective ... 

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