Fault Tree Analysis of an Industrial Accident

Fault Tree Analysis of an Industrial Accident
Accident Description
-425453721735Figure SEQ Figure \* ARABIC 1: Diagram of Three Mile Island Nuclear Power Plant 2Source: Nuclear Safety Analysis Center (1980)00Figure SEQ Figure \* ARABIC 1: Diagram of Three Mile Island Nuclear Power Plant 2Source: Nuclear Safety Analysis Center (1980)The Three Mile Island, Pennsylvania accident was a 1979 nuclear powerplant meltdown caused by a mechanical failure in the cooling system. The mechanical failure led to the core meltdown of the second reactor. Additionally, radioactive gas leaked through the walls. Fortunately, there were no injuries. The powerplant had two reactors, both using a pressured water system. The first reactor produced 880 megawatts, while the second produced 959 megawatts (Nuclear Safety Analysis Center, 1980). Interestingly, the second reactor was brand new at the time of the accident, while the first had been in use for five years. The mishap in unit 2 occurred at 4 am when the reactor was working at 97% power (Nuclear Safety Analysis Center, 1980). It comprised of a glitch in the secondary cooling circuit, which raised the temperature in the primary coolant. Resultantly, the reactor shut down within one second. However, a pilot-operated valve failed to close, leading to a large amount of essential coolant escaping further, causing the meltdown of the reactor.
The valve on the reactor cooling framework opened shortly after the shutdown. However, it did not shut down after 10 seconds. Staying open, it released crucial reactor coolant to the coolant channel tank (Nuclear Safety Analysis Center, 1980). The administrators believed the valve had closed since the computer signalling system showed it had. Nonetheless, they lacked an instrument showing the valve’s exact position. High-pressure infusion pumps pushed water into the reactor framework in response to the cooling water deficiency. Eventually, cooling water flooded the pressurizer, which kept it from boiling. Operators answered by lessening the progression of increasing water and finally shut down the pumps because of the feared pipe damage that would cause a rupture. Consequently, the reactor coolant evaporated, uncovering the core, which became hotter, damaging the fuel rods and releasing the radioactive substance into the water.
Fault Tree Analysis (FTA)
Components of the System
The macronuclear powerplant system is a pressurized water reactor system with the following components as represented in figure 1 above:
* The reactor building houses a pressurized relief tank, pressurized relief valve, block valve, safety valve, control rods, steam generator, reactor coolant pump, and the reactor core.
* The turbine building consists of the turbine, generator, condenser, condensate pump, and main water feeding pump.
* Cooling tower for cooling the flowing coolant water.
* Transformer.
In this accident, the top event was the nuclear core meltdown leading to radioactive material leakage and a lack of power production. Basic events in the turbine building include power circuits, main feed water pump, circulating water pump, condensate pump, and condenser. Additionally, basic events in the reactor building include a pressurized relief tank, pressurized relief valve, block valve, safety valve, and reactor coolant pump.
-42530171893Basic Events1 – Circulating water pump failure (0.04)2 – Main feeder water pump failure (0.04)3 – Condensate pump failure (0.04)4 – Valve failure (0.05)5 – Duct Leakage (0.002)6 – Pressurized tank leakage (0.002)7 – Reactor coolant pump failure (0.05)8 – Pilot Valve Failure (0.05)9 – Pressurizer failure (0.003)Figure 2: Fault Tree Analysis Source: Author (2022)00Basic Events1 – Circulating water pump failure (0.04)2 – Main feeder water pump failure (0.04)3 – Condensate pump failure (0.04)4 – Valve failure (0.05)5 – Duct Leakage (0.002)6 – Pressurized tank leakage (0.002)7 – Reactor coolant pump failure (0.05)8 – Pilot Valve Failure (0.05)9 – Pressurizer failure (0.003)Figure 2: Fault Tree Analysis Source: Author (2022)Fault Tree Analysis
Probability Calculation and Quantification
According to the International Atomic Energy Agency (2005), nuclear energy stations are among the safest and most secure globally. Yet, mishaps can occur, antagonistically influencing individuals and the climate. To limit the probability of a mishap, the International Atomic Energy Agency helps the Member States apply global wellbeing principles to reinforce thermal energy station security. The primary failure in nuclear power plant systems is the cooling system. The liquid coolants consist of deuterium, water, liquified sodium, and polyphony. Europe, the USA and different nations, and the International Atomic Energy Agency have laid out a few fundamental objectives and necessities for future thermal energy stations.
In Europe, the significant utilities have cooperated to propose a typical arrangement of atomic security necessities known as the European Utility Requirements (International Atomic Energy Agency, 2005). The objective is to lay out a strict framework of utility necessities in Europe to permit the advancement of cutthroat, normalized plans that would be licensable in the separate nations. Additionally, the United States Division of Energy has sent off a significant worldwide exploration drive named Generation IV to create and show better reactor advances than ever (World Nuclear Association, 2022). Wholesome objectives have been laid out as a feature of the Generation-IV exertion. Client prerequisites reports have likewise been arranged in Japan, the Republic of Korea and China.
After the Three Mile Island incident, the Nuclear Safety Analysis Centre (1980) in America did a detailed accident report. Among the quantifiable probabilities based on the report review and logical deductions, the above basic events can be quantified in probability as follows: Circulating water pump failure (0.04), a main feeder water pump failure (0.04), condensate pump failure (0.04), valve failure (0.05, duct Leakage (0.002), pressurized tank leakage (0.002), reactor coolant pump failure (0.05), pilot valve failure (0.05), pressurizer failure (0.003). Probability for the top event (nuclear core meltdown) leading to no power production = Cooling Failure x Core Exposure.
Pnuclear core meltdown = (Pcooling failure) (Pcore exposure)
First, calculating cooling failure probability
Pcooling failure = Psecondary cooling circuit failure + Pcoolant drain
Psecondary cooling circuit failure = Pcirculating water pump failure (0.04) + Pmain feeder water pump failure (0.04) + Pcondensate pump failure (0.04)
Psecondary cooling circuit failure = 0.12
Pcoolant drain = Pvalve failure (0.05) + Pduct leakage (0.002)
Pcoolant drain = 0.052
Pcooling failure = 0.12 + 0.052
Pcooling failure = 0.172
Second, calculating core exposure probability.
Pcore exposure = (Pprimary cooling circuit failure) (Pdestruction of control rods)
Pprimary cooling circuit failure = Ppressurized tank leakage (0.002) + Preactor coolant pump failure (0.05)
Pprimary cooling...