Fault trees reactors

Despite its shortcomings, WASH-1400 provides at this time (1978) the most complete single picture of accident probabilities associated with nuclear reactors. The fault tree/event tree approach coupled with an adequate database is the best method available to quantify these probahililies. 

Fig. 3.4.S-3 Fault Trees at the Nodes of Event Tr< t- to Determine the Probability. RPS is reactor protection .y.ii m SIS is safety injection sy.stem...

It is illustrated in Section 3.4.4 by tracing the paths for leaking engine compression and applied to fault tree construction for the FFTF reactor Fullwood and Erdmann, 1974). The method involves writing Boolean equations for all paths whereby hazardous material may be released. It is primarily useful for enumerating release paths, but not for what started the release It was used to enumerate the possible paths for stealing nuclear bomb material from a facility. 

The QRA was conducted by risk sts and design innel to determine the probability of explosive releases of the chemical. Fault tree analysis identified several combinations of equipment failures and operator errors that could cause the top event (reactor explosion), Failure data were obtained from plant ex ice and industry da%.ui,/uoes to quantify the fault trees to estimate the frequency of reactor explosions. The fault trees suggested several safety improv-... 

Anticipated Transients without Scram for Light Water Reactors, Vol. 1-3, December 1978. Haasl, D. F, et al., Fault Tree Handbook, January 1981. 

Crosetti, P. A., 1971, Fault Tree Analysis for Reactor Systems, Inst. Power Ind. 14 p 54. 

Fullwood, R. and R. C. Erdman, 1974, On the Use of Leak Path Analysis in Fault Tree Construction for Fast Reactor Safety, CONF-74040I-P3. 

As can be noted in Figure 21.7.2, steam and ediane are mi.xed before entering die reactor tubes where pyrolysis reacdons take place. All feed and product lines must be equipped with appropriate control devices to ensure safe operation. The FTA flow chart breaks down a TOP event (see descripdon of fault tree in Unit II) into all possible basic causes. Aldiough, diis mediod is more structured than a PHA, it addresses only one individual event at a dine. To use an FTA for a complete liazard analysis, all possible TOP events must be identified and investigated this would be extremely time consuming and perhaps urmecessary in a preliminary design. 

Another widely-used predictive method is the use of Fault Tree Analysis. This is a reversethinking method. The analyst assumes an accident or specific undesirable event—the so-called TOP Event. This could be the release of a toxic gas from a reactor safety relief valve. 

Fio. 26. The industrial implementation of a fault tree for the reactor-quench example. 

NUCLEAR HAZARD CATEGORY 1 FACILITIES. Fault tree/eveut tree analyses are required if the facility is a large reactor. If the facility is not a reactor and a PSM Rule PrHA is required, the analyses can be conducted as described for Nuclear Hazard Category 2 Facilities. Different systems or processes within the facility may be analyzed using different methods. For example, HAZOP studies may be used as the PrHA method for processes that contain chemical hazards. Fault tree/event tree analyses may be used to analyze systems that do not need to comply with the PSM Rule. 

Toward the end of the Second World War, systems techniques such as fault tree analysis were introduced in order to predict the reliability and performance of military airplanes and missiles. The use of such techniques led to the formalization of the concept of probabilistic risk assessment (PRA). The publication of the Reactor Safety Study (NRC, 1975)—often referred to as the Rasmussen Report after the name of principal author, or by its subtitle WASH 1400—demonstrated the use of such techniques in the fledgling nuclear power business. Although WASH 1400 has since been supplanted by more advanced analysis techniques, the report was groundbreaking in its approach to system safety. 

An Intermediate Event is an event that is one that has been designated for further development as the analysis progresses. For example, when developing the fault tree to do with the loss of production just discussed, an intermediate event could be Reactor shuts down. It is likely that, as the tree is developed, this simple phrase will be expanded as various reasons for reactor shutdown are considered. The reasons for reactor shutdown could include loss of feed, loss of flow of heating medium, or instrument malfunction. 

Fig. 9.47 Fault tree for an automatic reactor trip system (without activation)...

As seen from the fault tree of Fig. 9.47, where the activation of the reactor trip (denoted by SB in Fig. 9.46) was not included for avoiding too much complexity), the automatic reactor trip can fail for the following reasons ... 

Figure 9.57 shows the system for injecting an inhibitor into a reactor. The corresponding fault tree is presented in Fig. 9.58. The system mainly consists of the injector vessel containing the inhibitor, the corresponding measuring devices and valves, and a catch tank. In case the temperature is too high temperature switch TSH opens valve AV5 and the inhibitor is injected into the reactor by a pressure blanket inside the injector vessel. Redundantly, pressure switch PI opens valve...

A standard reactor used in the process industry for synthesis reactions is shown in Fig. 9.59. The corresponding fault tree model is presented in Fig. 9.60. Reactants A and B are introduced in controlled quantities into the reactor. A catalyst is continuously supplied and the temperature as well as the pressure increases are measured. The protective trip system consists of the safety valve SVl and the relief system made up of pressure switch PSHHl, relay 1, and pneumatic valve AVI. The safe place for relief is considered to be a discharge tank just as that of Fig. 4.11, which is modelled as in Fig. 9.56. 

Fig. 9.60 Fault tree for the pressure relief trip system of the reactor of Fig. 9.59...

Kazan has two significant difficulties. One is that for some failure modes, including those involving human error, data is lacking or uncertain. The other problem is that it is hard to estimate the effectiveness of mitigatory measures, particularly those taken by the process operator. Lees has collected information on the failure rates of various types of valves, thermocouples, pressure switches and other instruments. The same author and his associates - have reviewed and analysed data from incidents in batch reactors and constructed fault trees from the data on incident frequency. 

Figure 5.2 Fault trees for reactor overpressure outline generic tree.

The RPS is a very important system for NPP safety. So that, current regulation requires a very high reliability of the system in performing its vital and required safety function, i.e. reactor trip. The fault tree shown in Fig. 3 can be used to estimate the reliability of the RPS before and after the STl change under study. Table 1 shows there are no significant reduction of the RPS reliability after the change, since it remains very high. 

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