Fault trees (FT) and event trees (ET) are widely used in industry to model and evaluate the reliability of safety systems. This work seeks to analyze and estimate the core damage frequency (CDF) due to flow blockage (FB) and loss of coolant accident (LOCA) due to large rupture of primary circuit pipe with respect to a specific 10 MW Water-Water Research Reactor in Ghana using the FT and ET technique. Using FT, the following reactor safety systems: reactor protection system, primary heat removal system, isolation of the reactor pool, emergency core cooling system (ECCS), natural circulation heat removal, and isolation of the containment were evaluated for their dependability. The probabilistic safety assessment (PSA) Level 1 was conducted using a commercial computational tool, system analysis program for practical coherent reliability assessment (SAPHIRE) 7.0. The frequency of an accident resulting in severe core damage for the internal initiating event was estimated to be 2.51
Following the Fukushima accident, the International Scientific Community’s attention to unmitigated nuclear power plant (NPP) accidents and their mitigation has become a major highlight [
Yazdi, Hafezi, and Abbassi in 2019 demonstrated the relevance and viability of the probabilistic safety analysis methodology in a high-tech industry. The study also compared its results with common conventional approaches [
This paper uses the probabilistic safety analysis (PSA) technique to examine the probability of occurrence of an accident and evaluate its consequences, providing a numerical estimate to indicate how safe the water-water reactor (VVR) 10 MWth is. The PSA analysis is also employed in approximating the risk cutback that could be accomplished by making alterations to the nuclear reactor design or the operation and maintenance practices.
The series of occurrences that could result in fuel integrity loss or core damage and their frequency of occurrence were identified and quantified in PSA Level 1. This paper presents the results obtained as part of the probabilistic safety assessment (PSA) Level 1 conducted for the 10 MWth VVR research reactor. Accident-causing initiating events (flow blockage and loss of primary coolant) were identified, described, and evaluated.
The event involving the blockage of one channel only has the greatest probability of occurrence with the possibility of resulting in the meltdown of fuel elements. From the initiating events in the loss of primary coolant category, large loss of coolant accident (LOCA) is the one that causes the largest consequences due to the possibility of uncovering the core in less time than all other events in the category. It was for the abovementioned reasons that analysis of the sequences of the accident for the two initiating events was selected. The selection criteria took cognizance of the event with the highest frequency/probability of occurrence and the one that results in more severe consequences for the reactor core and the events analyzed were loss of flow due to blocked channel and loss of coolant accident (LOCA) due to large rupture.
The paper also focused on the estimation of core damage frequency in the 10 MWth VVR in the eventuality of the selected initiating events. The progression of the accident and performance of the systems to mitigate each event were analyzed employing the event tree (ET). Furthermore, the reliability of these systems was quantified using the fault tree (FT).
Moreover, there are only two reactors in the West African subregion. The size and nature of the reactors do not warrant PSA since they are inherently safe and have only a few IEs that could be prescribed. The PSA was performed to find severe accident weaknesses and provide quantitative results to support the decision-making for 10 MW VVR based on Russia and Ghana intergovernmental proposal for NPP and research reactor. Again, the skills and expertise acquired from this work are useful to the Nuclear Power Program Ghana is pursuing. This work will also aid in the identification of safety issues and cost-effective solutions to safety problems which arise during the design of 10 MW VVR. Moreover, it will also provide valuable guidance to the areas in which additional funds available for improvement of the overall safety of the reactor can most effectively be spent. Regardless of all innovations performed by the new generation of reactors, the likelihood of accidents and faults in the safety systems remains. Consequently, PSA is performed to examine the reliability of reactors concerning design basis accidents, which considers the likelihood of damage to the reactor core in the most different accident scenarios [
The water-water reactor (VVR) is a pool-type reactor. The research reactor has a thermal power of 10 MWth. It has a light water coolant and moderator together with a Beryllium reflector surrounding the active core. It is 19.7% enriched with a peak neutron flux of 1014 cm−2·s−1. It has eleven horizontal channels (with each channel being a double open-ended channel with a diameter of 150 mm). The reactor contains fifty sample irradiation positions. A critical facility with a peak neutron flux of 107 cm−2·s−1 is also constructed to function as a mock-up of the reactor. The cylindrical reactor vessel which contains the core has an inner diameter of 0.652 m and a height of approximately 2.8 m. It is located off-centre in a shielded central tank with the same height as the central tank with an outer diameter of 2.3 m. The nominal flow rate is 1250 m3/h. The water flows downward under forced convection induced by the primary cooling pumps. The water flows upwards from beneath the central tank through a single 0.35 m diameter pipe. It flows further up and is directed towards the centre and trickles down through the core. It finally exits the core from the centre of the reactor vessel through a single 0.35 m diameter pipe. There is no other piping penetration in the reactor vessel and central tank. A cross section of the 10 MWth research reactor is shown in Figure
Research nuclear reactor (VVR): 1–tank; 2–fuel assembly (FA); 3–beryllium reflector; 4–reactor vessel; 5–7–baffle, guiding, and support grids; 8, 9–suction and pressure pipelines [
Block diagram of the reactor cooling system and connected system.
This section outlines the summary of the concepts of probability associated with the uncertainty analysis used in the PSA. The discussion shall present the basic concepts and principles as used in the SAPHIRE 7.0 code [
MCS
A common approach to calculating the probability for a top event is to add together the probabilities for the cutsets where the cutset probability is given by (
This paper focuses on Level 1 PSA studies and seeks to use the knowledge acquired to establish and consolidate methodologies for future reliability studies. The following plant operation states were considered: nominal full-power operation (10 MWth); reduced power operation; start-up operation; reactor subcritical, reactor pool availability. From the safety point of view, when a plant operating state brackets all others, it is referred to as a nominal full-power operation. This is attributed to the fact that the reactor pool constitutes a large heat sink that is always available, regardless of the operating state of the reactor. Figure
Flow chart for the methodology.
The accident-causing initiating events, loss of coolant accidents (LOCA) and flow blockage (FB), were chosen for this study. LOCA consists of every single event that directly generates loss of integrity of the primary coolant pressure boundary whiles FB is the entrance blockage of the primary coolant through the fuel element channel or channels.
The design of the 10 MWth VVR incorporates five basic safety functions. The safety functions are aimed at preventing core damage following an initiating event. Event tree for IE loss of coolant accident (LOCA) (ET1) was used to model the reply of the reactor to coolant loss. In this work, the sequence of events resulting in coolant accident loss was detected and modelled in SAPHIRE. The safety systems and safety functions that could sequentially happen to mitigate the LOCA are briefly described as follows.
It was supposed that, in the case of reactor usage at full-power, there was a decapitate break of the biggest (12 inches) pipe joined to the lowest part of the reactor culminating in the outset of the succession of events.
The reactor protection system (RPS) was mechanized and labor-intensive systems shut down the reactor after LOCA. The accomplishment of this event led to scram and therefore stoppage of the fission chain reaction.
The pool was cut off from the system for cooling in the event of a malfunction of the RPS. This happened when the butterfly valves were shut, either physically or mechanically, around only some minutes after the accident. Effective pool isolation from the break location leads to the core being deepened in the pool water. Therefore, if reactor pool isolation (RPI) is achieved, coolant supply continues to be accessible (coolant loss curtails). The interest currently was if natural circulation of the pool water was accessible.
During natural circulation heat removal (NCHR), the flapper valve opened when the pool water was available and this enabled natural cooling of the core. Natural cooling was enough to avoid core impairment because there was effective scram. Therefore, sequence number 1 was successful. If the flapper had failed to unlock to allow natural circulation, sufficient heat elimination would have been unlikely. It was expected that core impairment would happen. Similarly unlikely was the sufficiency of heat elimination with or without natural heat removal competency in case the reactor was not closed through the RPS. Once more it was expected that if a shutdown was not realized, after complete loss of flow, core damage would happen. A distinct amount of core inventory discharge was expected contingently on whether the fission chain reaction was halted and on whether natural circulation was likely.
The handiness of natural pool water circulation was immaterial if the pool was not isolated, and the emergency cooling system starts functioning to spray water on the core. It was supposed that, with the reactor shutdown, core spraying was enough to take away decay heat and consequently circumvent core damage. Therefore, sequence number 5 was deemed a successful sequence. If the emergency core cooling system (ECCS) had failed, natural cooling in just air would not have been enough to take away decay heat and therefore core damage will happen and radioactivity will be discharged inside the reactor building. Outcomes though are contingent on the radioactivity quantity discharged out of doors of the containment and therefore the ensuing events necessarily must be involved in the tree. Given that ECCS was not working, then isolation of the reactor building was likely, and every gate and door stay shut. This action has the objective of having to stop any discharge of radioactivity into the surroundings. Effective containment isolation (CI) needs an emergency ventilation system to operate to release the pressure and take away the maximum amount of the discharged radioactivity through the filters. The severest accident sequence, from the discharge viewpoint, develops when the containment is unsuccessful at isolating as it signifies the discharge of the largest amounts of radioactivity from the entire accident sequences (accident sequences number 8 and number 17).
In case ECCS is unsuccessful, NEV (LOCA) was employed to model the emergency ventilation systems operation in the event of LOCA. This system working effectively signified retaining most of the radioactivity in the containment and the filters. System failure (accident sequences number 7 and number 16) signified larger amounts of radioactivity discharge when the system functions but lesser than when the containment isolation (CI) failed (sequence number 8 and number 17). This event tree determined seventeen (17) sequences. Two sequences culminated in a safe state (sequence numbers 1 and 5). All other sequences (sequence number 2, 3, 4, 6 to 17 as shown in Figure
Event tree (ET1) for the initiating event LOCA.
Event tree for IE flow blockage (FB) (ET2) was used to model the reaction of the reactor system to an event culminating in blockage of the coolant flow in one of the channels. Moreover, the reaction that separated this event tree from the former ones is the reaction of the RPS, the irrelevance of the primary heat removal system, as well as the emergency ventilation system. The events comprising the event tree (ET2) are flow blockage (the IE), reactor protection system (RPS), primary heat removal, natural circulation heat removal, containment isolation, and emergency ventilation system.
With the flow blockage event tree (ET4), the assumption made was that there was blockage of a fuel coolant channel, in which case there was an incomplete loss of flow leading to an accident condition. When the flow blockage accident scenario was initiated, only manual scram as a RPS was available to manually shutdown the reactor because there were no mechanized signals for either shutdown or reverse.
Because of the failure of the RPS to mitigate the accident scenario, there was a need to isolate the reactor building. In the event of the failure of the reactor building to isolate, the emergency ventilation system was applied to model the functioning of the emergency ventilation system (NEV) in the event of ‘flow blockage’. Again, only manual initiation was possible. ET4 consisted of four event sequences as shown in Figure
Event tree (ET2) in the event of flow blockage.
Quantitative risk (a systematic approach for evaluating likelihoods, consequences, and risk of adverse events) analysis based on event and fault tree used in this work employed two basic assumptions; the first assumption was related to likelihood values of input events, whilst the other assumption was concerning interrelation among basic events. Traditionally, event trees and fault trees both use probabilities; however, to tackle the issue of uncertainties, the assumption of probability distributions of input event likelihoods is employed [
The event tree was used to define the initiating event within the reference research reactor. It was then employed in analyzing the course of events that follow as determined by the operation or failure of the safety systems provided to prevent the core from melting and to stop the release of radioactivity to the surroundings [
PSA involves several analytical methods. These include the development of event tree and fault tree logic models used for the analysis of accident sequences. Large break LOCA (loss of coolant accident) was chosen as an initiating event for application in this paper. The assumption made for the chosen initiating event was that during full-power operation; there is a double-ended rupture of the largest (12 in) pipe connected to the base of the reactor [ LOCA (initiate event (IE)): it is assumed that during full-power operation, there is a double-ended rupture of the largest (12 inches) pipe connected to the base of the reactor. Availability of reactor protection system: following LOCA, the reactor protection system, both automatic and manual systems, should shut down the reactor. The success of this event results in scram and hence in interruption of the fission chain reaction. Pool isolation: following LOCA, the pool should be isolated from the cooling system.
This occurs if the butterfly valves close, either manually or automatically, within 16 min following the accident. Successful isolation of the pool from the location of the break results in the core being immersed in the pool. The probability of failure is generally less than 0.1 and therefore the probability of success is always close to 1. Thus, the probability associated with the upper (success) branches in the tree is assumed to be 1 [
The frequency of occurrence in a sequence of events is the product of the conditional probabilities of the individual events in that chain. In this study, if the successive events in a sequence are independent, then the frequency of a sequence was the product of unconditional probabilities of the individual events (so each front-line system has P failures as identical) [
(a) System fault tree with top event “reactor protection system failure” in LOCA. (b) System fault tree for “no signal” with top event “reactor protection system failure” in LOCA.
System fault tree with the top event “no containment isolation”.
System fault tree with top event “emergency core cooling system”.
(a) System fault tree with the top event “no emergency ventilation.” (b) System fault tree for “emergency ventilation does not start” with the top event “no emergency ventilation”.
System fault tree with top event “natural circulation heat removal”.
(a)Ssystem fault tree with the top event “no reactor pool isolation.” (b) System fault tree for “butterfly valve fails to close” with the top event “no reactor pool isolation”.
Fault tree with top event “primary heat removal”.
The entrance blockage of the primary coolant through the fuel element channel or channels can occur when objects unintentionally fall on the reactor core which results in the coolant flow reduction. The coolant flow reduction can cause a local overheating of the fuel element plate followed by failure of the cladding. The flow blockage (FB) can be detected by the operator, visual inspection during the operation, a significant increase in pressure loss in the core, measured by pressure transducer located at the top of the pool (corresponding to a value above 10% of nominal flow), a significant increase in coolant temperature at the out of the core, or the radiation detectors positioned below the movable platform supporting the core in the worst-case scenario.
There are no available resources in the reactor to allow the automatic detection of the flow blockage when few channels are blocked because in this situation the detectors of differential pressure and temperature increase at the out of the coolant system cannot detect small variations. If there is no detection by the operators (visually), the reactor will not be shut down and may cause local damage to the fuel element plates.
Where there is a deterioration of fuel element plates cooling, where channels are blocked, it can lead the plates to their melt. In this case, there will be the release of fission products to the pool water and the atmosphere of the containment, with its detection by radiation monitors and automatic shutdown of the reactor through the protection system. The regular ventilation system will be off and the emergency ventilation system will be activated and thereby isolating the containment area. The emergency ventilation system, therefore, pushes the air to the filters decreasing the release of radioactive material to the environment. To mitigate a blocked channel, the following are necessary safety functions so that there is no destruction of the core, and the release of radioactivity to the surroundings does not come to be above the permissible limits in shutdown of the reactor by the operator on the reactor protection system; maintenance of the containment is through turning off the regular ventilation system and activation of the emergency and isolation ventilation system.
The expected sequences of events, in this case, are blockage of few cooling channels of a fuel element caused by some object, without the possibility of automatic detection; visual detection of the event by the operators and manual shutdown of the reactor; turning off the regular exhaustion and insufflating the containment area, and the start of the emergency exhaustion of this area; and isolation of the containment area.
Figure
The frequency of occurrence of the channel blockage IE was obtained per year. The Greek reactor frequency is equal to 10−2/year [
Using the SAPHIRE program [
The estimated CDF value can be considered OK and acceptable because this type of accident would cause only minor damage to the core which would not cause large releases of radionuclides to the environment. It is only local damage to few fuel plates and the VVR-10 reactor has containment and systems that mitigate potential releases of radiation above the limits permissible for the population. Furthermore, when this result is compared with other research reactors, it is of the identical order of magnitude as the Greek reactor [
Both the flow blockage and the large break LOCA could lead to damage to the core and this conclusion is based on the consideration that the establishment of natural circulation of coolant through the core would be enough to mitigate the circumstance of the other initiating events, removing the residual heat of it. However, this depends on the decoupling of the convection valve that can fail. If there is no establishment of natural circulation, there might be damage to the reactor core for some of the initiating events described. Considering the failure in establishing the natural circulation, the most critical situation would be the initiating event of locking the pump shaft, because the flywheel would not act and the forced circulation would be interrupted, and consequently there would be a greater amount of residual heat to be removed.
The studies presented in this paper considered a large break LOCA and channel flow blockage of the 10 MW VVR research reactor. We have used the fault of the system tree approach to determine the top event probabilities in each system, i.e., reactor protection system (RPS), pool isolation (PI), natural circulation heat removal (NCHR), emergency core cooling system (ECCS), containment isolation (CI), and emergency ventilation (EV). Applying the values of the probabilities assigned to each basic event in each front-line system as shown in Tables Total frequencies of beyond design basis accident sequences (plant damage states) are typically in the range of 1.0 × 10−5–1.0 × 10−4/plant and year. Evaluated higher frequencies are seen as indicators for safety improvements. Total frequencies of accident sequences with a potential of early and high activity releases caused by bypassing of the containment shall be of one order of magnitude lower than the abovementioned frequency values.
Description of basic events and probabilities for the reactor protection system (RPS).
Basic event | Components failure identification | Probability |
---|---|---|
1 | Failure of all 8 rods | 2.280 × 10−6 |
2 | Failure of electromagnets to engage | 2.920 × 10−6 |
3 | NOR no electric power | 9.999 × 10−1 |
4 | Gate slow scram fails | 6.970 × 10−4 |
5 | Relay T3 stuck closed | 6.970 × 10−4 |
6 | Sensor fails to give a signal | 2.350 × 10−2 |
7 | Relay T1 fails to open | 6.970 × 10−4 |
8 | No electric power | 2.790 × 10−6 |
9 | Human error | 1.000 × 10−2 |
10 | Sensor T2 fails | 2.350 × 10−2 |
11 | Relay T2 fails | 6.970 × 10−4 |
12 | Sensor fails to give a signal | 2.350 × 10−2 |
System fault tree with top event “containment isolation”.
Basic event | Components failure identification | Probability |
---|---|---|
1 | Gates of the ventilation system fails | 1.010 × 10−4 |
2 | Doors fail to remain closed | 1.440 × 10−5 |
3 | Air pump # 1 fails to stop | 3.240 × 10−4 |
4 | Air pump # 2 fails to stop | 3.240 × 10−4 |
5 | Air pump # 1 fails to stop | 3.240 × 10−4 |
System fault tree with the top event “no pool isolation”.
Basic event | Components failure identification | Probability |
---|---|---|
1 | Operator fails to close manual butterfly valves | 1.000 × 10−2 |
2 | Manual butterfly valves fail in an open position | 3.600 × 10−6 |
3 | Operator fails to give a signal to pneumatic valves | 1.000 × 10−2 |
4 | No electric power | 2.790 × 10−6 |
5 | Pneumatic valves fail stuck in the open position | 8.200 × 10−5 |
6 | Human error | 1.000 × 10−2 |
7 | Sensor T2 fails | 2.350 × 10−2 |
8 | Relay T2 fails to open | 6.970 × 10−4 |
9 | Sensor fails to give a signal | 2.350 × 10−2 |
10 | Relay T1 fails to open | 6.970 × 10−4 |
System fault tree with the top event “no emergency ventilation.”
Basic event | Components failure identification | Probability |
---|---|---|
1 | No signal | 8.300 × 10−4 |
2 | Human error | 1.000 × 10−2 |
3 | Filters fail | 3.440 × 10−4 |
4 | Air pump | 2.270 × 10−3 |
5 | Two valves (h) fail to open | 1.680 × 10−3 |
6 | Loss of normal offsite power | 1.000 × 10−4 |
7 | AC generator fails | 2.790 × 10−3 |
8 | Diesel motor fails | 8.200 × 10−3 |
9 | Switches fail stuck | 3.200 × 10−2 |
System fault tree with top event “natural circulation heat removal failure.”
Basic event | Components failure identification | Probability |
---|---|---|
1 | Flapper fails to open (stuck) | 1.440 × 10−5 |
2 | Wrong weight | 1.000 × 10−2 |
System fault tree with top event “emergency core cooling system failure.”
Basic event | Components failure identification | Probability |
---|---|---|
1 | Hole in hose | 1.200 × 10−4 |
2 | Operator fails to connect hose | 1.000 × 10−2 |
3 | Water valves fail stuck closed | 3.600 × 10−6 |
4 | No water in the tank | 1.020 × 10−4 |
All accident sequences leading to core damage frequency.
Even tree | Frequency of releases (SAPHIRE 7.0) | Sequences |
---|---|---|
LOCA (ET1) | 2.510 × 10−4 | #2–#4, #6–#17 |
FB (ET2) | 1.450 × 10−4 | #2–#4 |
TOTAL | 3.960 × 10−4 |
Frequencies contributing most to core damage (100%).
Event tree | Frequency of releases (SAPHIRE 7.0) | Sequences |
---|---|---|
ET1 | 1.035 × 10−5 | #6 |
ET1 | 1.195 × 10−9 | #8 |
ET1 | 3.430 × 10−5 | #15 |
ET1 | 3.972 × 10−9 | #17 |
TOTAL | 4.470 × 10−5 |
Frequencies contributing most to core damage (50%).
Event tree | Frequency of releases (SAPHIRE 7.0) | Sequences |
---|---|---|
ET1 | 2.987 × 10−5 | #12 |
ET1 | 2.566 × 10−7 | #13 |
ET1 | 2.454 × 10−9 | #14 |
TOTAL | 2.013 × 10−5 |
Frequencies contributing most to core damage (30%).
Event tree | Frequency of releases (SAPHIRE 7.0) | Sequences |
---|---|---|
ET1 | 2.979 × 10−6 | #9 |
ET1 | 2.562 × 10−5 | #10 |
ET1 | 3.449 × 10−7 | #11 |
TOTAL | 2.894 × 10−5 |
Frequencies contributing most to core damage (10%).
Event tree | Frequency of releases (SAPHIRE 7.0) | Sequences |
---|---|---|
ET1 | 1.001 × 10−4 | #2 |
ET1 | 4.544 × 10−7 | #3 |
ET1 | 1.156 × 10−9 | #4 |
TOTAL | 1.467 × 10−5 |
Frequencies contributing most to core damage (<5%).
Event tree | Frequency of releases (SAPHIRE 7.0) | Sequences |
---|---|---|
ET2 | 1.000 × 10−4 | #2 |
ET2 | 4.400 × 10−5 | #3 |
ET2 | 1.160 × 10−6 | #4 |
TOTAL | 1.450 × 10−4 |
Furthermore, the results of Table
Comparison of large LOCA frequencies in various PSAs.
PSA | Frequency of releases |
---|---|
NUREG 4550-PWR [ | 5.000 × 10−4 |
RSS-PWR [ | 2.700 × 10−4 |
PSA-GRR [ | 1.200 × 10−4 |
The SAPHIRE computed the minimal cutset upper bound (top event) for the fault trees and a set of random samples from the uncertainty distribution of the basic events. SAPHIRE then computed the first few moments of the distribution and the mean, and 95th percentile values among others. The moments were calculated as a basis for comparison of the calculated distribution with other distributions from the first few moments, the sample mean and sample variance. The set of all possible executions of all the basic events (BE) in the model for the work was regarded as a population and any subset of those possible executions as a sample. A sample of values for one of the model response variables drawn from the population of all possible executions of the model was obtained by executing the model and recording the values, and the sample mean and the standard deviation were calculated from the sample values. From the sample statistics and a confirmed assumption about the distribution of the population’s response variable values, it was possible to calculate a confidence interval for the mean value of the response variable for the population, i.e., for all possible executions of the model. But it is worth noting that such confidence interval was for the model (the population of all possible model executions), not for the research reactor. Nevertheless, the conventional validation interpretation of the confidence interval is that if the observed research reactor value for the response variable was included in the model confidence interval, then the model was considered valid (or not invalid) for that response variable. There is no statistical justification or refutation for this interpretation. Statistically, the calculated confidence interval relates to the population of possible model executions. Table
Uncertainty values for the fault trees.
Fault tree | FREQ/Year | Mean | Median | 95TH | Minimum | Maximum | Std Dev |
---|---|---|---|---|---|---|---|
CI | 1.440 × 10−5 | 1.440 × 10−5 | 1.440 × 10−5 | 1.440 × 10−5 | 1.440 × 10−5 | 1.440 × 10−5 | 0.000 |
ECCS | 1.024 × 10−2 | 1.024 × 10−2 | 1.024 × 10−2 | 1.024 × 10−2 | 1.024 × 10−2 | 1.024 × 10−2 | 0.000 |
NCHR | 1.001 × 10−2 | 1.001 × 10−2 | 1.001 × 10−2 | 1.001 × 10−2 | 1.001 × 10−2 | 1.001 × 10−2 | 2.248 × 10−9 |
NEV (FB) | 4.291 × 10−3 | 4.291 × 10−3 | 4.291 × 10−3 | 4.291 × 10−3 | 4.291 × 10−3 | 4.291 × 10−3 | 0.000 |
NEV (LOCA) | 4.531 × 10−3 | 4.531 × 10−3 | 4.531 × 10−3 | 4.531 × 10−3 | 4.531 × 10−3 | 4.531 × 10−3 | 1.384 × 10−9 |
PHR | 4.848 × 10−4 | 4.848 × 10−4 | 4.848 × 10−4 | 4.848 × 10−4 | 4.848 × 10−4 | 4.848 × 10−4 | 4.930 × 10−11 |
RPI | 1.010 × 10−3 | 1.010 × 10−3 | 1.010 × 10−3 | 1.010 × 10−3 | 1.010 × 10−3 | 1.010 × 10−3 | 2.974 × 10−10 |
RPS (LOCA) | 3.286 × 10−3 | 3.286 × 10−3 | 3.286 × 10−3 | 3.286 × 10−3 | 3.286 × 10−3 | 3.286 × 10−3 | 0.000 |
For each of the columns in Table
The confidence intervals were calculated using the
For the 5000-sample size taken, the 95% confidence interval computed for each sample means that 95% of the intervals contained the population mean. Naturally, 5% of the intervals did not contain the population mean.
Consequently, the larger confidence levels made it more likely that the research reactors response variable was inside the model confidence interval for that variable and thus results in a less rigorous validation test [
The core damage frequency (CDF) obtained for the channel blockage was 1.45 × 10−4/year, which can be considered to be within acceptable limits and so would cause only minor local damage to the core and consequently without large releases of radionuclides to the environment. Comparing the estimated frequency with some research reactors, the result obtained can be considered satisfactory, as it is of the identical order of magnitude as the Greek research reactor [
The paper only considered level 1 PSA; however, level 2 PSA must be applied to the reference reactor studied in this work to model sequences of accident leading to accident progression phenomena analysis and estimating the frequency of different accident release categories (large early radioactivity releases to the environment). Furthermore, developing PSA Level 3 for the reference 10 MW Russian research reactor to model the potential impact on the environment assessed based on offsite accident management measures, population distribution, and predominant meteorological conditions will enhance the safety of the reactor. Another issue worth analyzing would be to investigate whether binary decision diagrams (BDDs) allow fault trees and event trees in performing PSA, thus allowing better estimates of components at the same time conserving proportionate computational time.
Alternating current
Binary decision diagrams
Core damage frequency
Event tree
Fuel assembly
Flow blockage
Fault tree
Greek research reactor
Initiating event
International atomic energy agency
Loss of coolant accident
Minimum cutset
Nuclear power plant
Nuclear Regulatory Guide 4550
Probabilistic safety analysis
Pressurized water reactor
Reactor protection system
Reactor safety study (reported in WASH-1400)
Software Application Programme for Hands-on Integrated Reliability Evaluation
Water-water reactor
The report of the Rasmussen study that effectively started the use of probabilistic safety assessment.
The raw data supporting the conclusions of this research will be made available by the authors without undue reservation.
The authors declare that they have no conflicts of interest.