Postulated accident sequences

In the different reactor risk studies performed up to now, a number of potential accident sequences have been analyzed in detail. The conditions prevailing in the different stages of such an accident have been described in numerous publications, with regard to KWU-designed plants, for example, by Hassmann et al. (1987). These physical and chemical conditions (such as temperature, pressure, flow characteristics, hydrogen-to-steam ratio) vary widely between the different postulated accident sequences, influencing the release and transport behavior of the fission products. In the following sections only those parameters will be discussed which are of importance for fission product chemistry and behavior. 

The different investigated accident sequences can be subdivided into two main categories  

It has to be pointed out again that such initiating events will only lead to a severe core damage if all the emergency core cooling systems failed to operate. If only a part of them (they are installed on a conservative basis with sufficient overcapacity and redundancy, see Sections 1.1.4. and 2.1.4.) functions, the postulated accident sequence would be terminated before severe core degradation occurred. This is particularly true for plants with extensive accident management systems. 

The details of the core melt-down processes depend both on the particular composition of the respective reactor core and on the particular conditions of the accident sequence under consideration. However, as a comparative evaluation of the results from different experiments on core melt progression performed in test reactors and from examination of the damaged core of the TMI-2 reactor (Hob-bins et al., 1991) has shown, in spite of considerable differences in the conditions and physical scales, several important phenomena seem to be common to all such events. Eutectic interactions between core materials cause the formation of liquid phases and loss of original core geometry at comparatively low temperatures (about 1500 K). The first liquids to be formed are metallic in nature and consist 

This process of melting down of the reactor core inside the reactor pressure vessel is associated with an extensive release of the volatile fission products and a partial volatilization of the other fission products, uranium and the actinides (invessel release). The volatilized fuel constituents are transported by the steam-hydrogen flow out of the core region, together with volatilized fractions of the core structural materials (stainless steels, Ni alloys). 

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Safe shutdown of the reactor due to the passive strong reactivity feedbacks without motion of control rods either due to operator or automatic scram action in response to all credible and postulated accident sequences ... 

The minimal core height is 66 cm, with a diameter of 358 cm, and includes a total of 192 fuel assemblies (Table 9.7). The small sodium void worth could be reduced by further pancaking the core, which would come at the expense of increases in the bumup reactivity loss. Initial evaluations of postulated accident sequences indicated attractive safety characteristics for the burner core, comparable to a conventional (breakeven) core design. Since the reactivity insertion from withdrawal of control rods is larger for the pancaked core, smaller incremental movements (rod stop adjustments) are required. 

The heat transport from the core melt into the sump water, as well as the production of permanent gases in the core melt - concrete interaction, will result in a pressure increase inside the closed containment. Provided that no measures for controlled depressurization are undertaken (see next section), then, according to the results of thermodynamic calculations, the steam-gas pressure within the containment would reach the postulated failure value of the containment steel shell of about 0.9 MPa after 5 to 10 days, depending on the accident sequence. Such an overpressure failure will not be a catastrophic burst of the shell, but rather the enlargement of the operational leaks to a size permitting the escape of gas and steam at a rate high enough to keep the pressure inside the containment at a... 

Heinrich states that the occurrence of an injury accident invariably results from a completed sequence of factors culminating in the accident itself. He postulates five factors or stages in the accident sequence, with the injury invariably caused by the accident, and the accident in turn the result of the factor that immediately precedes it. 

Malfunctioning of power operated valves could create unacceptable situations during accidents and contributes to the risk associated with postulated core-melt accident sequences. 

Late containment failure is defined as eontainment failure postulated to occur later than 24 hours after the onset of core damage. Sinee the probabilistic risk assessment assumes the dynamic phenomena, such as hydrogen combustion, to occur before 24 hours, this feilure mode occurs only from the loss of containment heat removal via failure of the passive containment cooling system. The fission products that are airborne at the time of containment failure will be discharged at high pressure to the environment, as the containment blows down. Subsequent release of fission products can then pass directly to the environment. Accident sequences with failure of containment heat removal are grouped in release category CFL. 

A containment isolation failure occurs because of the postulated failure of the system or valves that close the penetrations between the containment and the environment. Containment isolation failure occurs before the onset of core damage. For such a failure, fission-product releases fi-om the reactor coolant system can leak directly from the containment to the environment with diminished potential for attenuation. Most isolation failures occur at a penetration that cormects the containment with the auxiliary building. The auxiliary building may provide additional attenuation of aerosol fission-product releases. However, this decontamination is not credited in the containment isolation failure cases. Accident sequences in which the contairunent does not isolate prior to core damage are grouped into release category Cl. 

Pre-operational testing is performed as system and equipment availability allows. The interdependence of systems is also considered. Sequencing of the start-up tests depends on specified power and flow conditions and intersystem prerequisites. The start-up test schedule establishes that, prior to core load, the test requirements are met for those plant stmctures, systems and components that are relied upon to prevent, limit or mitigate the consequences of postulated accidents. Commissioning is sequenced so that the safety of the plant is not dependent on untested systems, components or features. 

Accident sequences initiated by external events (postulated initiating... 

Accident sequences initiated by postulated initiating events with a significant probability of being contemporaneous with external events, even though they are not correlated. 

If there is no residual water in the lower plenum, a possibility for some accident sequences, the melt would directly attack the lower head (see Section 3.5.2). However, the progression treated here postulates the more complex case in which residual water exists in the lower plenum. 

In Reference 13-25, BWX Technologies (BWXT) provided a reactor core and reactor module assembly sequence that ensured the system remained within the nuclear criticality safety license requirements during all phases of manufacture and test, including postulated accident conditions. This information had not been submitted to NR nor factored into the assembly and testing flow process at the time of project cancellation. 

While the maximum potential consequences are similar for the process spill and the SCB fire DBAs, the process spill scenario dominates the overall risk to the public by several orders of magnitude as compared to all other DBAs. The maximum potential consequence at the exclusion area boundary (3000 m.) is calculated to be 2 Rem, and the sequence of events that results in this dose is assessed to be extremely unlikely. Accidents that are expected to occur within the lifetime of the facility are (conservatively) calculated to have dose consequences of up to several miliirem at the exclusion area boundary. These DBAs bound ail other potential accidents that have been postulated to occur in the facility. 

The low-pressure scenario which is initiated by a large-sized break in the primary circuit, that is an event similar to that of the loss-of-coolant design basis accident described in Section 6.2.1.. In the severe accident scenario it is additionally postulated that, after the action of the accumulators and the borated water storage tanks, the sump water recirculation pumps will fail to operate. Thus, the decay heat cannot be removed from the reactor core vnth the consequence that the water volume present inside the reactor pressure vessel (RPV) begins to boil off at about atmospheric pressure. The AB sequence of WASH-1400 describes such a large-break scenario. In this low-pressure scenario, the treatment of fission product behavior inside the primary circuit is comparatively simple the probability of occurrence of such an accident, however, is extremely small. 

The containment event tree is a tool that provides a logical and practical stmcture for uniting the complex phenomenology of postulated severe accident event sequences. The treatment of severe accidents provided by the containment event tree provides assuranee that important contributors to fission-product release are identified and evaluated in a stmctured and diseiplined approach. The bases for the nodes on the tree are supported by analyses, evaluations and testing, empirical data from past studies, and by the APIOOO design. 

Postulated initiating events are occurrences that may lead to reactor fault sequences or accident scenarios. They originate from component failures, system malfunctions, human error or external events and special internal events. 

ETA is a system safety analysis technique for identifying and evaluating the sequence of events in a potential accident scenario following the occurrence of a postulated IE. ETA utilizes a visual logic tree structure known as an ET. 

The V-sequence refers to a Loss of Coolant Accident (LOCA) through the Low Pressure Coolant Injection System (LPIS), which in this type of nuclear power plants has a common part with the Residual Heat Removal System (RHRS). The suction pipe of the RHRS connects two of the three hot legs in the primary circuit with the RHRS pumps placed into the Auxiliary Building. It has been assumed that three isolation valves fail in one of these pipelines (two of them are motorized valves) a break near the RHR pump has been postulated as a result of the over-pressurization generated in the pipe. 

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