Failure Of Safety Systems Research Articles

A technique to generate plausible counter-operation procedures for an emergency situation based on a model expressing functions of components

ABSTRACTIn large-scale engineering plants like nuclear power plants, various anomalies and accidents are usually predicted and efficient measures to minimize the damage from them have been implemented. However, when an unexpected anomaly or accident occurs, operators are asked to take flexible responsive actions to avoid its development to a catastrophic situation and to minimize the influence and damage by the accident. Considering the lessons learned from Fukushima Daiichi accidents, it is important to develop a system to support counter activities by operators and plant staff in an accidental plant condition that is not covered by emergency operating procedures prepared. The purpose of this study is to develop a technique for generating plausible operation procedures under such unexpected emergency situations. The multilevel flow modeling (MFM), a method to represent the functions of components and their relations, is applied as a framework to model functions of components. The technique uses some additional data and knowledge for deriving operation procedures. The applicability of the technique is confirmed by trial operation procedure generations for pressurized water reactor (PWR) loss-of-coolant accident cases with the conditions of some failures of safety systems.

Reliability prediction of seismic switch for early detection of earthquakes around NPPs

The recent Fukushima accident has posed several challenges to nuclear community with respect to safety of the nuclear power plants (NPPs) during simultaneous occurrence of seismic, Tsunami, multiple failures of safety systems and also maintaining the long term sub criticality during these events. Safety Instrumented Systems (SISs) are used in many application areas to reduce the likelihood of such severe accidents and/or to mitigate the consequences. An indigenously developed seismic switch is employed to safely shut down the reactor upon the occurrence of a seismic event. The reliability of seismic switch is of vital importance to ensure safe operation of a NPP. This paper presents the results of reliability study performed on digital and analog seismic switches to assess their performance.

Characteristics of debris in the lower head of a BWR in different severe accident scenarios

Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy.The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20tons) and large (>100tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input can result in large changes in the final mass of debris. Typical ranges of the debris properties in different scenarios are discussed in detail.

Integration of functional reliability analysis and system hardware reliability through Monte Carlo simulation

Passive safety systems are being increasingly used in innovative reactor designs due to its inherent safety characteristics. The passive safety system failure can occur due to process uncertainties and changes in system hardware state with time. The failure probability contribution due to process uncertainties of a passive decay heat removal system of a fast breeder reactor are quantified through functional reliability analysis. The probability density of the peak hot pool temperature obtained from functional reliability analysis is combined with system hardware reliability through Monte Carlo simulation. The probabilities of crossing the various design safety limits on hot pool temperature are evaluated. The failure probabilities are estimated for different standard deviation values of the peak hot pool temperature probability density. This leads to a linear relationship between failure probability and standard deviation having two distinct regions. The results from the present analysis are compared with the results from two other methods. The results from the three methods converge when the process uncertainty contribution to total failure probability is dominant. The results are significantly different from the three methods when the process uncertainty contribution to total failure probability is not dominant. The difference or the convergence of results by different methods is governed by the extent of process uncertainty contribution to the total failure probability. The proposed method reduces the conservatism as compared to the functional reliability analysis combined with fault tree approach.

Simulation of Containment Pressurization in a Large Break-Loss of Coolant Accident Using Single-Cell and Multicell Models and CONTAIN Code

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code’s results.

The Investigation of Nonavailability of Passive Safety Systems Effects on Small Break LOCA Sequence in AP1000 Using RELAP5 MOD 4.0

The passive safety systems of AP1000 are designed to operate automatically at desired set-points. However, the unavailability or failure to operate of any of the passive safety systems will change the accident sequence and may affect reactor safety. The analysis in this study is based on some hypothetical scenarios, in which the passive safety system failure is considered during the loss of coolant accidents. Four different cases are assumed, that is, with all passive systems, without actuation of one of the accumulators, without actuation of ADS stages 1–3, and without actuation of ADS stage 4. The actuation of all safety systems at their actuation set-points provides adequate core cooling by injecting sufficient water inventory into reactor core. The LOCA with actuation of one of the accumulators cause early actuation of ADS and IRWST. In case of LOCA without ADS stages 1–3, the primary system depressurization is relatively slow and mixture level above core active region drops much earlier than IRWST actuation. The accident without ADS stage 4 actuation results in slow depressurization and mixture level above core active region drops earlier than IRWST injection. Moreover, the comparison of cladding surface temperature is performed in all cases considered in this work.

Post-test analysis of the ROSA/LSTF and PKL counterpart test

Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety.Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper presents the results obtained in the simulations for both installations using TRACE, observing, in general, a good agreement with the experiments. However, ROSA/LSTF calculations underestimated the maximum PCT value, what might be explained by the higher core level predicted in the simulation compared with the experiment. In PKL calculations, PCT maximum value is slightly higher than in the experiment, and the core level predicted is lower. In the comparison of the evolution of both installations a different timing in the transient events is observed, due to the difference in the pressure vessel design. Thus, when PKL vessel is modified with some of the ROSA/LSTF features, the evolution of the new PKL model behaviour is closer the one observed in ROSA/LSTF calculations.

Safewash! Risk attenuation and the (Mis)reporting of corporate safety performance to investors

This paper responds to the recent call in Safety Science for greater academic attention to voluntary corporate disclosure of occupational health and safety information. The study extends prior literature by focusing in detail on occupational injury performance measures provided publicly by Australia’s 50 largest listed firms. Content analysis of a sample of 239 annual reports issued between 1997 and 2009 revealed, as expected, firms in hazardous industries provided significantly higher rates of disclosure, in terms of both number of performance measures and frequency of reporting, than firms in less hazardous industries. Yet higher disclosure did not translate to adequate reporting on measures needed to demonstrate the prevention of serious injury. Instead, there was a strong reliance on variations of highly aggregated frequency rates often criticised in academic and practitioner safety literature. Further, evidence appeared to demonstrate strategic efforts to reduce the visibility of high-consequence safety system failures over time, thereby attenuating investor perceptions of occupational safety risk. The results provide important insights into the communication of corporate OHS performance to investors and suggest the presence of quantitative performance measures is not necessarily indicative of high quality disclosure. The findings also have important implications for the development of disclosure indices in future voluntary reporting research.

Probabilistic risk analysis of safety management system failure and impact on economic performance: the case of jet engine manufacturing

It is widely known that operational accidents affects the productivity of manufacturing industry and evidences show that the adoption of a safety management system (SMS) can influence economic performance. This paper presents a method to perform probabilistic risk analysis of safety system failure and analysis of the impact on costs and economic performance. The novelty is the use of probabilistic risk analysis with Bayesian belief network to evaluate the effects of safety system failure on costs. The method is applied in the overhaul of jet engines to identify the factors affecting the safety system failure, the probability of their occurrence and the impact on cost. This method utilises design thinking to classify the risks. This study adds to the growing understanding that probabilistic risk analysis of safety system failure and associated costs is essential for safety operation and for competitive performance in jet engine manufacturing system.

Investigations on in-vessel melt retention by external cooling for a generic VVER-1000 reactor

External or internal hazards, combined with multiple failures of components and safety systems or human errors can lead to a reactor core melt. In that case the reactor pressure vessel is the last barrier to keep the molten materials inside the reactor and to prevent further challenges to the nuclear power plant structures and consequently to the environment. In-vessel melt retention by external vessel cooling is a possible mitigative severe accident measure. Up to the moment it is not considered as a severe accident management strategy for VVER-1000 reactors. In this paper we analyse the possibility of in-vessel melt retention for a generic pressurized water VVER-1000 reactor during the late in-vessel phase of a postulated station blackout scenario.We developed a numerical model describing the thermal behaviour of a segregated molten pool situated in the lower plenum of the reactor pressure vessel and the thermo-mechanic behaviour of the vessel wall. The finite element code ANSYS® was used for the simulations.The results show that the highest thermo-mechanical loads are observed in the vertical part of the vessel wall, which is in contact with the molten metal. Parameter studies on the thickness of the metal layer have also been performed. Without flooding, the vessel wall will fail, as the necessary temperature for a balanced heat release from the external surface via radiation is near to or above the melting point of the steel. However, the external flooding could help the retention of the corium within the reactor pressure vessel.

Medication Errors: Scope and prevention strategies

Background: Medication errors are a significant public health concern. Although significant advances have been made, errors are still relatively common and represent an opportunity for healthcare improvement.Methodology/Principal Findings: Since the publication of To Err is Human, medication errors have been under tremendous scrutiny. Organizations have moved towards a non-punitive approach to evaluating errors. This approach to medication errors has aided in identifying common pathways to medication errors and improving understanding regarding the anatomy of a medication error. As a result, prevention strategies have been developed to target common themes contributing to errors. Error prevention strategies may target common contributors of medication errors, broadly grouped as performance lapses, lack of knowledge, and lack or failure of safety systems. Strategies to thwart medication errors range from process improvement to integration of technology in the health care environment.Conclusions/Significance: Organizations should devote resources to address medication error prevention strategies in an effort to improve patient outcomes and decrease morbidity and mortality associated with medication errors.

Preliminary evaluation of effect of Engineered Safety Features on source term for AHWR containment

Abstract The proposed Advanced Heavy Water Reactor (AHWR) employs double containment envelope along with many Engineered Safety Features (ESFs) to mitigate the consequences of Loss-of-Coolant Accidents (LOCA) with safety system failure, during which high enthalpy steam and radioactive fission products will be discharged into the containment. In such conditions, the pressurized containment will be the source of activity release to the environment by way of leakage. It is required to study the effect of ESFs on the source term from the AHWR containment. An analysis was performed to evaluate the release rate from the AHWR containment during a postulated accident with the in-house containment code CONTRAN and the aerosol behavior code NAUA5-M in a coupled way. Modules for simulating the engineered safety features were incorporated in the CONTRAN code and the aerosol transport behaviour was evaluated using NAUA5-M separately. The AHWR containment is divided into three nodal volumes interconnected by junctions. The blow down mass, energy discharge data and activity released into the containment from the reactor core, for a postulated LOCA case of 200% RIH break with failure of shutdown systems (1 &amp; 2), are inputs to the CONTRAN code. Thermodynamic parameters like containment gas temperature, partial pressure of steam, air in the subdivided volumes along with the flow rates through junctions obtained from CONTRAN were supplied to NAUA5-M. An analysis was carried out for a number of cases, postulated based on availability/unavailability of ESFs. Pressure, temperature and activity concentration transients were evaluated, for 72 h, in the subdivided volumes along with the activity released out of the containment through leakages and stack discharges for all the cases. This paper highlights the importance of operation of ESF in reducing the activity release to the environment.

Assessment methodology for confidence in safety margin for large break loss of coolant accident sequences

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis (BEPU) is increasingly being used for deterministic calculation in NPPs. The PSA methodology integrates information about the postulated accident, plant design, operating practices, component reliability and human behavior. The deterministic and probabilistic methodologies are combined by analyzing the accident sequences within design basis in the event trees of a postulated initiating event (PIE) by BEPU. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with calculating the safety margin with 95% confidence and 95% probability in large break loss of coolant accident (LBLOCA). In the present study, five uncertain input parameters were selected. Uniform probability density function was assigned to the uncertain parameters in the selected range and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for five parameters was randomly mixed by LHS to obtain 25 input sets. An event tree for the initiating event, LBLOCA inside containment, has been used from a VVER study on Level-1 PSA and four non-core damage (NCD) accident sequences were identified for this study. In the accident analysis the success and failure of safety systems reflected in event tree was appropriately modeled in the system thermal hydraulics code runs. The PCT was obtained for each of 25 code runs for each accident sequence. A Kolmogorov – Smirnov goodness-of-fit test carried out for PCTs indicated that they followed normal distribution for each of the accident sequences. The probability distribution of safety margin (difference between acceptable value and PCT) in each accident sequence was also obtained. The values of safety margin for the 95% confidence and 95% probability are estimated. The robustness of the system can be judged based on this. This paper describes the methodology. LBLOCA in a VVER type reactor is considered as an example.

Environmental Health Insights into the 2011 Tōhoku Japan Earth quake Disaster.

Environmental Health Insights into the 2011 Tōhoku Japan Earth quake Disaster.

Formal Considerations in Establishing Risk-Consistent Acceptance Criteria for Reactor Safety

Traditionally, the safety of a nuclear reactor system has been assessed through a set of mechanistic calculations of bounding accident sequences using conservative models. Extensive experience in the operation and analysis of nuclear reactor systems has led to two complementary approaches: best-estimate mechanistic calculations with a quantitative estimate of the uncertainty for assessing conformance with acceptance criteria based on technical limits and probabilistic risk analysis of the event sequences due to the probability of failure of safety systems. Both assess the safety of the reactor system; however, the emphasis, especially in the estimation of probabilities, is different in the two approaches. Yet both address the same concern: the safety of the reactor system. We discuss the formal relations that are necessary for a risk-consistent analysis of the safety of the nuclear reactor systems with respect to the two current approaches.

A model for medication safety event detection

Hospital medication safety event detection predominantly emphasizes the identification of preventable adverse drug events (ADEs) through self-reports. These relatively rare events only provide insight into patient harm and self-reports identify only a small portion of ADEs. A broader system-focused approach to medication safety event detection that uses an array of event detection methods is recommended. This approach illuminates medication system deficits and supports improvement strategies that can prevent future patient risk. To: (i) describe a system-focused approach to hospital medication safety event detection, and (ii) present a case illustration of approach application. SYSTEM-FOCUSED MODEL AND METHODOLOGY: A three-level medication safety event detection model that ranges from a narrow harm-focused to broader system-focused approach is described. A standardized cross-level methodology to detect medication safety events is presented. A Level 3 system-focused methodology that incorporated both voluntary and non-voluntary event detection strategies was used in 17 critical care (n = 4), intermediate care (n = 7) and medical-surgical units (n = 6) across two hospitals. A total of 431 events were detected: 78 (18.1%) ADEs and 353 (81.9%) potential ADEs. Of the 353 PADEs, 302 (70.0%) were non-intercepted events. Non-voluntary detection methods yielded the majority of events (367, 85.1%). The incidence of ADEs was low when compared with non-intercepted PADEs. This was indicative of medication safety system failures that placed patients at risk for potential harm. Non-voluntary detection methods were much more effective at detecting events than traditional self-report methods.

A Near Miss: A Nitrous Oxide-Carbon Dioxide Mix-up Despite Current Safety Standards

A Near Miss: A Nitrous Oxide-Carbon Dioxide Mix-up Despite Current Safety Standards

Dynamic risk assessment using failure assessment and Bayesian theory

To ensure the safety of a process system, engineers use different methods to identify the potential hazards that may cause severe consequences. One of the most popular methods used is quantitative risk assessment (QRA) which quantifies the risk associated with a particular process activity. One of QRA's major disadvantages is its inability to update risk during the life of a process. As the process operates, abnormal events will result in incidents and near misses. These events are often called accident precursors. A conventional QRA process is unable to use the accident precursor information to revise the risk profile. To overcome this, a methodology has been proposed based on the work of Meel and Seider (2006). Similar to Meel and Seider (2006) work, this methodology uses Bayesian theory to update the likelihood of the event occurrence and also failure probability of the safety system. In this paper the proposed methodology is outlined and its application is demonstrated using a simple case study. First, potential accident scenarios are identified and represented in terms of an event tree, next, using the event tree and available failure data end-state probabilities are estimated. Subsequently, using the available accident precursor data, safety system failure likelihood and event tree end-state probabilities are revised. The methodology has been simulated using deterministic (point value) as well as probabilistic approach. This Methodology is applied to a case study demonstrating a storage tank containing highly hazardous chemicals. The comparison between conventional QRA and the results from dynamic failure assessment approach shows the significant deviation in system failure frequency throughout the life time of the process unit.

Decay Heat Removal in GEN IV Gas‐Cooled Fast Reactors

The safety goal of the current designs of advanced high‐temperature thermal gas‐cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct‐cycle gas‐cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5‐3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

Bayesian approaches for detecting significant deterioration

Risk indicators can provide useful input to risk management processes and are given increased attention in the Norwegian petroleum industry. Examples include indicators expressing the proportion of test failures of safety and barrier systems. Such indicators give valuable information about the performance of the systems and provide a basis for trend evaluations. Early warning of a possible deterioration is essential due to the importance of the systems in focus, but what should be the basis for the warning criterion? This paper presents and discusses several Bayesian approaches for the establishment of a warning criterion to disclose significant deterioration. The Norwegian petroleum industry is the starting point for this paper, but the study is relevant for other application areas as well.