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One of the lessons learned from the Fukushima Daiichi accident is the emergency operating procedures (EOPs) have to be improved. The BWR Owners’ Group revised the emergency procedure guidelines and addressed the lesson learned from the Fukushima Daiichi accident in revision 3 in order to avoid loss of turbine-driven makeup water systems during reactor depressurization. However, the improvement deserves much more attention. The existing EOPs at the time of the accident may not be adequate enough for the prolonged station blackout condition, because resources required for performing the EOPs are vastly unavailable or gradually exhausted. The improved EOPs must not only permit early reactor pressure vessel depressurization, but also address the risk accompanied with the emergency depressurization. For this reason, Taiwan Power Company proposed the Ultimate Response Guideline (URG) to cope with Fukushima-like accidents. The main content of the URG is a two-stage depressurization strategy, namely the controlled depressurization and the emergency depressurization. The technical basis of the two-stage depressurization strategy was discussed in this paper. The effectiveness of the URG was verified by using TRAC/RELAP Advanced Computational Engine (TRACE). Besides, the emergency responses performed by Fukushima Daini nuclear power plant (Fukushima Daini NPP) were found to be very similar to the URG. The consequences of Fukushima Daini NPP somehow demonstrate that the URG is effective for Fukushima-like situation. Effective strategies for coping with Fukushima-like accidents are worthwhile to be integrated into the improved EOPs. The Fukushima Daiichi accident might have been prevented if the improved EOPs had been used, even with the existing design of Fukushima Daiichi NPP at the time of the accident.

Loss Of Offsite Power (LOOP) accident is one of the most important Design Basis Accidents (DBAs) that originate due to loss of the house-load demands which make serious concerns for each NPP. Moreover, in the case of loss of all diesel generators (due to earthquake, etc.), the accident forwards to the Station Black-Out(SBO). Following and learning the Fukushima accident, the cares have been focused on using internal and/or external water resources for the passive management of the SBO accident. In the current study, passive strategies in the VVER-1000/V446 NPP are proposed using a tank in the secondary loop (called Deaerator) for the SBO management. After RELAP5 modeling benchmarking in the steady-state as well as standard SBO management (which are mentioned in the NPP FSAR), two operational SBO management are proposed herein including “Feeds and Bleeds” (in the primary and secondary loops) and also including a Deaerator tank in the secondary side. The results show that using the Deaerator-water tank of the secondary side as well as feed and bleed, the time available for operators before the core heat-up (i.e. fuel cladding temperature of 1200 °C); extends to 35000 s. Obviously; it gives a good opportunity for plant operators to access further time for using other facilities/methods to avoid core damage.

In an emergency condition of nuclear power plant, operators have to mitigate the accident in order to remove the decay heat and to prevent the release of radioactive material to the environment following the emergency operating procedures (EOPs). The action of operators on a component, for example, changing the parameter level of a component, which is described in a procedure step, will impact other components of the plant and the plant behavior. Nowadays, the advanced main control rooms have been equipped with the computer-based procedures (CBPs) which provide some features and benefits which are not available in paper-based procedures. However, most of CBPs do not provide information of the impact of the counteractions on each procedure step (components influenced and future plant behavior) although it is useful for operators to understand the purpose of the procedure steps before making decisions and taking the actions. This paper discusses the functional information and the method to generate the information using multilevel flow modeling (MFM) model of operator actions on some procedure steps of a simplified EOP of pressurized water reactor (PWR) plant, as an example.

When an accident occurs, operators in nuclear power plants (NPPs) must follow emergency operating procedures (EOPs) or severe accident management guidelines (SAMGs). However, EOPs and SAMGs are sym...

This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating state with unsealed primary circuit by removing the MCP head. The reference nuclear power plant is Unit 6 at Kozloduy NPP (KNPP) site. RELAP5/ MOD3.2 computer code has been used to simulate the transient for WWER-1000/V320 NPP model. A model of WWER-1000 based on Unit 6 of KNPP has been developed for the RELAP5/MOD3.2 code at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS), Sofia. The plant modifications performed in frame of modernization program have been taken into account for the investigated conditions for the unsealed primary circuit. The most specific in this analysis compared to the analyses of NPP accidents at full power is the unavailability of some important safety systems. For the purpose of the present investigation two scenarios have been studied, involving a different number of safety systems with and without operator actions. The selected initiating event and scenarios are used in support of analytical validation of Emergency Operating Procedures (EOP) at low power and they are based on the suggestions of leading KNPP experts and are important in support of analytical validation of EOP at low power.

The objective of this paper is to present the current organization of the Emergency Procedures including Emergency Operating Procedures (EOP) and Severe Accident Management Guidelines (SAMG) in Kozloduy Nuclear Power Plant (KNPP) as a function of the severity of the accident conditions. Special attention is paid to SAMG. It is described when the SAMG are used and at which conditions in a transition between the EOPs and the SAMG should be made. The Critical Safety Function Restoration Guidelines and their connections with SAMGs and EOPs are also discussed. The arrangement of SAMG is described in detail, since in the KNPP exist 2 types of SAMGs for Main Control Room (MCR) and for the Accident Management Centre (AMC) and they contain the same strategies, but they are different in format. Both types are symptom oriented procedures, but those for MCR are in 2-column-format with interconnections, whereas those for the AMC are developed in a logical manner and simplified for people, who take decisions. In the paper, they are also discussed the adopted strategies in existing SAMG that should be followed to recover from a damaged core condition and to prevent or mitigate the release of fission products. In the paper, they are also described a number of technical measures for management and mitigation of severe accidents, which are implemented in KNPP before and after the Fukushima accident. Many of them are common for WWER-1000 type of reactors, but some of them are unique and plant specific. This information can be useful for operators of other WWER type reactors or even PWR reactors.

This paper presents the results of analytical validation of operator actions in case of “Steam Generator Tube Rupture” (SGTR) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The purpose of the analyses is to demonstrate the ability to terminate primary to secondary leakage and to indicate an effective strategy for preventing secondary leakage to the environment and in this way to prevent radiological release to the environment. Following depressurization and cooldown of reactor coolant system (RCS) with isolation of the affected steam generator (SG), in these analyses are validated options for post-SGTR cooldown by: • back up filling the ruptured SG; • using letdown system in the affected SG and • by opening Fast Acting Isolation Valve (FAIV) and using Steam Dump Facility to the Condenser (BRU-K). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The RELAP5/MOD3.2 computer code has been used for the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS). This paper is possible through the participation of leading specialists from KNPP.

During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 ( Carlson et al., 1990 ). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model ( Fletcher and Schultz, 1995 ). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

This paper presents the results of thermal-hydraulic calculation of “Steam Line Break” analysis at full power reactor for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of symptom based emergency operating procedures (SB EOPs) for this plant. The RELAP5/MOD 3.2 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the systems thermal-hydraulics code RELAP5/MOD 3.2 at the Institute for Nuclear Research and Nuclear Energy–Bulgarian Academy of Sciences (INRNE–BAS), Sofia. The main purpose of the analysis is to estimate the parameters of the monitored plant which are used to identify symptoms that are used by operators to identify the plant's state and the critical safety function (CSF). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The performed analysis is based on a previously used bounding approach in analytical validation of SB EOPs. Based on this approach a list of scenarios has been performed, involving a different number of safety systems with or without operator actions. The presented thermal-hydraulic calculations of the accident scenarios involve the loss of CSF “Subcriticality” for VVER-1000/V320 units at KNPP.

Emergency operating procedures (EOPs) are fundamental for coping with emergency and accident situations in nuclear power plants. Researchers at the Halden Reactor Project have developed a design co...

This paper presents the results of thermal-hydraulic calculation of accident scenarios that involve the trip of main coolant pump (MCP) in case of Small break loss of coolant accident (SB LOCA) for WWER-440/V230 units at Kozloduy Nuclear Power Plant (KNPP), done in support of the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The main purpose of these analyses is to show how the different time of MCP switching off results in primary inventory depletion in case of SB LOCA and it is reflect on peak cladding temperature. According to this, the SB LOCA scenario is regarded from the point of view of an inadequate core cooling. Therefore, the primary concern is Critical Safety Function (CSF) “Core cooling” and “Primary inventory”. High core residual heat, minimal safety injection flow and other initial conditions challenging the mentioned CSFs are the main particularities of the accepted scenarios. The RELAP5/MOD3.2 computer code has been used to perform the analyses in a WWER-440 Nuclear Power Plant (NPP) model. A model of WWER-440 based on Unit 4 of Kozloduy NPP has been developed for the system’s thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia.

Computer-based emergency operating procedures offer some benefits compared with the paper-based EOP. However, most of them do not fully provide clear and necessary information related to the procedure steps, such as future plant behavior and the affected components after taking an action of each step because of the limited space available on the virtual display unit. This paper investigates a technique to derive and indicate such kind of information using a Multilevel Flow Modeling model of simplified EOP of steam generator tube rupture accident of a pressurized water reactor plant. In this case, the MFM model is built based on operator’s action on each procedure step. Then, by using the influence propagation rules for function primitives of MFM model, the influenced functions of plant components necessary for confirming the procedure step can be gathered. This information will help operators to understand and take the actions in order to achieve the objective of each procedure step.

Nuclear Power Plant (NPP) operators complete multiple types of tasks within Emergency Operating Procedures (EOPs). Due to the potential serious consequences of committing an error, it is important to determine if the workload (WL) demands operators encounter are at acceptable levels. This study investigates whether there are workload differences are distinct between task types and if there is a difference between each task type over multiple sessions in a simulated environment. Previous research supports that EEG, ECG, and the NASA-TLX are sensitive to changes in WL. The present preliminary experiment sought to investigate WL changes for experienced participants over a number of sessions and task types. During each session, participants completed tasks derived from a combination of EOPs and subject matter expert input that consisted of checking, detection, and response implementation task types. WL changes were measured through EEG, ECG, and NASA-TLX responses. The results indicate that WL differences were found among the different task types, but not sessions. The implications for these findings are discussed in detail.

Parameters and concepts are introduced in order to provide a view of nuclear thermal hydraulics covering from the bases to the application of the discipline. Namely, selected cross-cutting issues from the thermal hydraulics universe are defined. This view is essential for understanding the content of the book and for self-confirming the level of understanding of thermal hydraulics by the reader. Two phase flows and heat transfer are concerned together with safety concepts. Examples moving from bases to applications are the fundamental relationship of thermal hydraulics involving quality, void fraction, slip ratio and pressure, the steam binding phenomenon, the design basis accident, the defense in depth, and the final safety analysis report.

The Fukushima-Daiichi accident shook the world, as a well-known plant design, the General Electric BWR Mark I, was heavily damaged in the tsunami, which followed the Great Japanese Earthquake of 11 March 2011. Plant safety functions were lost and, as both AC and DC failed, manoeuvrability of the plants at the site virtually came to a full stop. The traditional system of Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG) failed to protect core and containment, and severe core damage resulted, followed by devastating hydrogen explosions and, finally, considerable radioactive releases. The root cause may not only have been that the design against tsunamis was incorrect, but that the defence against accidents in most power plants is based on traditional assumptions, such as Large Break LOCA as the limiting event, whereas there is no engineered design against severe accidents in most plants. Accidents beyond the licensed design basis have hardly been considered in the various designs, and if they were included, they often were not classified for their safety role, as most system safety classifications considered only design basis accidents. It is, hence, time to again consider the Design Basis Accident, and ask ourselves whether the time has not come to consider engineered safety functions to mitigate core damage accidents. Associated is a proper classification of those systems that do the job. Also associated are safety criteria, which so far are only related to ‘public health and safety’; in reality, nuclear accidents cause few casualties, but create immense economical and societal effects . for which there are no criteria to be met. Severe accidents create an environment far surpassing the imagination of those who developed EOPs and SAMG, most of which was developed after Three Mile Island . an accident where all was still in place, except the insight in the event was lost. It requires fundamental changes in our present safety approach and safety thinking and, hence, also in our EOPs and SAMG, in order to prevent future ‘Fukushimas’.

This paper presents the results of thermal–hydraulic calculation of accident scenarios that involve the loss of critical safety function (CSF) “Heat sink” for WWER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant at low power and cold condition. The main purpose of this analysis is to provide the response of monitored plant parameters to identify symptoms available to the operators and define timing for reaching the following stages during the development of processes in the reactor system: • Reaching the saturated temperature at the outlet of the assembly. • Beginning of reactor core uncovery. • Heating up of fuel. • Defining the transition time between EOPs and SAMG at temperature of 923 K. • Restoring of water level in the core. • Defining the CSF “Heat sink” status and the time of its loss. The results of the thermal–hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs at low power and cold condition. The principal acceptance criteria for EOPs are averting the onset of core damage. The RELAP5/MOD3.2 computer code has been used in performing the analyses in a WWER-1000 Nuclear Power Plant (NPP) model. A model of WWER-1000 based on Unit 6 of Kozloduy NPP has been developed for the systems thermal–hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy–Bulgarian Academy of Sciences (INRNE–BAS), Sofia. The low power and cold condition and the modifications after the modernization program are taken into account.

This paper presents the thermal–hydraulic investigation “Primary to secondary leakage at low power” for analytical validation of Symptom Based Emergency Operating Procedures (SBEOPs) at low power for VVER440/V230 units 3 and 4 at Kozloduy NPP. The main purpose of this analysis is to provide the response of monitored plant parameters to identify symptoms available to the operators and define timing for reaching the following stages during the development of processes in the reactor system, after primary to secondary leakage: • Reaching the saturated temperature at the outlet of the assembly. • Beginning of reactor core uncovery. • Define the time for reaching maximum level in the damaged SG. • Termination of natural circulation after decreasing of water level in reactor vessel below the hot nozzles. • Defining the time for opening SVs of damaged SG. • Define the CSF status and the time of loss of Critical Safety Functions. The thermal–hydraulic analysis for analytical validation of Symptom Based Emergency Operating Procedures (SBEOPs) at low power of reactor for VVER440/V230 units 3 and 4 at Kozloduy NPP is performed with RELAP5/MOD3.2 computer code. The RELAP5/MOD3.2 model of Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios have been developed and validated in the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE–BAS) Sofia, and Kozloduy NPP. The model provides a significant analytical capability for the specialists working in the field of NPP safety.

Gori-I nuclear power plant has been examining the effectiveness and efficiency of the current emergency operation procedures from human factors viewpoint. Previous study showed that some mistakes that the procedures did not comply with the writers` guidelines. Reviewing the current writers` guidelines for emergency operating procedures revealed that they lack of some important human factors rules such as enumeration of switching conditions and detailed action requirements, definite expression for setup points, description for anticipated results, and recommendation for use of present tense, affirmative sentence and active voice. This study suggested a new classification system for the writers` guideline contents supplementing the deficiencies of the current emergency operation procedure text.

With the development of information technology, computerized emergency operating procedures (EOPs) are taking the place of paper-based ones. However, ergonomics issues of computerized EOPs have not been studied adequately since the industrial practice is quite limited yet. This study examined the influence of step complexity and presentation style of EOPs on step performance. A simulated computerized EOP system was developed in two presentation styles: Style A: one- and two-dimensional flowcharts combination; Style B: two-dimensional flowchart and success logic tree combination. Step complexity was quantified by a complexity measure model based on an entropy concept. Forty subjects participated in the experiment of EOP execution using the simulated system. The results of data analysis on the experiment data indicate that step complexity and presentation style could significantly influence step performance (both step error rate and operation time). Regression models were also developed. The regression analysis results imply that operation time of a step could be well predicted by step complexity while step error rate could only partly predicted by it. The result of a questionnaire investigation implies that step error rate was influenced not only by the operation task itself but also by other human factors. These findings may be useful for the design and assessment of computerized EOPs.

In nuclear power plants, operators are allowed to follow EOPs (Emergency Operating Procedures) when reactor tripped because of accidents. But, it’s very difficult to diagnose accidents and find out appropriate procedures to mitigate current accidents in a given short time. Even if they diagnose accidents quickly, it also has possibility to misdiagnose. Methodology using Influence Diagrams has been developed and applied for representing the dependency behaviors and uncertain behaviors of complex systems. An example to diagnose the accidents such as SLOCA and SGTR with similar symptoms has been introduced. From the constructed model, operators could diagnose accidents at any states of accidents. Also, The integrated design of Thermal Hydraulics Online Monitoring Advisory System(THOMAS) is introduced in this paper. THOMAS, for improved Korea Standard Nuclear Power Plant, is an advisory monitoring system that uses the digital engineering technology such as virtual reality, and database. It is believed that th...

The appropriateness of the task complexity (TACOM) measure that can quantify the complexity of procedural tasks was validated in this study. To this end, two sets of task performance time data that have been extracted under the simulated steam generator tube rupture (SGTR) conditions of the reference nuclear power plant (NPP) A and B were compared with the associated TACOM scores. As a result, it was observed that two sets of task performance time data seem to be soundly explained by the associated TACOM scores. Although more additional activities should be conducted to clarify the appropriateness of the TACOM measure, the result of this study provides a crucial clue supporting that the complexity of emergency tasks stipulated in emergency operating procedures (EOPs) can be properly quantified by the TACOM measure.

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.

A verification study to evaluate the appropriateness of the diagnostic rules of the CE-type EOP for a SLOCA event was performed to address the diagnostic related ambiguities usually existed in behavioral characteristics of the plant systems in response to a specific initiator. A proposed method to verify the appropriateness of the diagnostic rules is to analyze the dynamic behaviors of the diagnostic parameters during the anticipated period of the operator's diagnosis time. The analysis of the dynamic behaviors is conducted by using the thermal-hydraulic (T/H) computer code, MARS, as the break size of the SLOCA event is increased (e.g., 0.76, 1.1, 1.4, 1.6, and 1.9 in.) under a realistic operative status of the related plant systems. According to the analysis results, two of the diagnostic rules revealed a mismatched condition with the simulated physical behaviors for the SLOCA event.

It is well known that the performance of human operators, which is one of the predominant factors affecting the safety of any human-involved systems, is largely attributable to the complexity of tasks. From this regard, in order to evaluate the complexity of tasks stipulated in procedures, Park developed the TACOM measure that consists of five sub-measures. In this study, in order to unravel a potential problem observed in the course of validation studies about the TACOM measure, the revised TACOM (R-TACOM) measure was suggested based on the theory of a task complexity. Although more detailed studies are necessary, the result of this study indicates that the R-TACOM measure is a more proper measure than the TACOM measure in quantifying the complexity of tasks stipulated in procedures.

It is well known that complicated procedures frequently cause human performance related problems that can result in a serious consequence. Unfortunately a systematic framework to evaluate the complexity of procedures is very rare. For this reason Park et al. suggested a measure called TACOM ( Ta sk Com plexity) which is able to quantify the complexity of tasks stipulated in procedures. In addition, it was observed that there is a significant correlation between averaged task performance time data and estimated TACOM scores. In this study, for an additional verification activity, TACOM scores are compared with operators’ performance data that are measured by Operator Performance Assessment System (OPAS). As a result, it is believed that TACOM scores seem to be meaningfully correlated with OPAS scores. Thus, it is reasonable to expect that the result of this study can be regarded as a supplementary evidence for supporting the fact that TACOM measure is applicable for quantifying the complexity of tasks to be done by operators.

In this study, the appropriateness of the task complexity (TACOM) measure that can quantify the complexity of emergency tasks was investigated by comparing subjective workload scores with the associated TACOM scores. To this end, based on the NASA-TLX (task load index) technique, 18 operators were asked to subjectively estimate perceived workload for 23 emergency tasks that were specified in the emergency operating procedures of the reference nuclear power plants. As the result of comparisons, it was observed that subjective workload scores increase in proportion to the increase of TACOM scores. Therefore, it is expect that the TACOM measure can be used as a serviceable method to quantify the complexity of emergency tasks

During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/V230 units at Kozloduy NPP, a number of analyses have been performed using the RELAP5/MOD3.2 computer code (RELAP5/MOD3 Code Manual. User's Guidelines. NUREG/CR-5535, INEL-95/0174, vols. 1–4, Rev. 1, 1990). Some of them discuss advantages and disadvantages of the use of Main Loop Isolating Valves (GZZs) in case of steam generator tube rupture (SGTR) transient. Use of the GZZs as a part of EOP mitigation strategies for SGTR accidents is analyzed in this paper to identify the behavior of important VVER parameters. A double-ended single-pipe break in SG #6 was chosen as representative. The results of these analyses demonstrate that sometimes GZZs – installed on the hot and cold legs of the primary loops – could provide safety function but sometimes their closing and re-opening could make the situation worse. RELAP5/MOD3.2 computer code has been used to simulate the SGTR transient in VVER 440 NPP model [RELAP5/MOD3 Code Manual. User's Guidelines. NUREG/CR-5535, INEL-95/0174, vol. 5, Rev. 1, 1995]. This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design bases scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

Many kinds of procedures have been used to reduce the operators' workload throughout various industries, such as in the aviation, the chemical and the nuclear industry. It is remarkable that, however, significant portion of accidents or incidents was caused by procedure related human error due to non-compliance of procedures. In this study, to investigate the operators' non-compliance behavior, emergency-training records were collected using a full scope simulator. And three types of the operators' behavior (such as strict adherence, skipping redundant actions and modifying action sequences) observed from collected emergency training records were compared with both their work experience and the complexity of procedural steps. As the results, three remarkable relationships are obtained. They are: (1) the operators who have an intermediate work experience seem to frequently adopt non-compliance behavior to conduct the procedural steps, (2) the operators seem to frequently adopt non-compliance behavior to conduct the procedural steps that have an intermediate procedural complexity, and (3) the senior reactor operators seem to accommodate their non-compliance behavior based on the complexity of procedural steps. Therefore, it is expected that these relationships can be used as meaningful clues not only to scrutinize the reason for non-compliance behavior but also to suggest appropriate remedies for the reduction of non-compliance behavior that can result in procedure related human error.

In this study, to suggest requisite characteristics for diagnosis procedures, the SROs' (senior reactor operators) behaviors in conducting the diagnosis procedure under simulated conditions are observed and analyzed in order to identify factors that make the performance of it difficult. As the results, four important characteristics that may be helpful in constructing effective and useful diagnosis procedures are suggested. These are: (1) the use of connection symbols should be minimized, (2) when connection symbols are used, precise information about entry symbols should be provided to the operators, (3) the use of negative expressions should be avoided, and (4) the diagnosis procedure should be constructed through the integration of both the non-compensatory and the compensatory approach. Although these characteristics are deduced, not from well-designed experiments, but from empirical observations, it is highly expected that the operators' diagnostic burden could be significantly reduced, if the above characteristics are properly embodied in the diagnosis procedures.

In complex systems, such as nuclear power plants (NPPs) or airplane control systems, human errors play a major role in many accidents. Therefore, to prevent an occurrence of accidents or to ensure system safety, extensive effort has been made to identify significant factors that can cause human errors. According to related studies, written manuals or operating procedures are revealed as one of the most important factors, and the understandability is pointed out as one of the major reasons for procedure-related human errors. Many qualitative checklists are suggested to evaluate emergency operating procedures (EOPs) of NPPs. However, since qualitative evaluations using checklists have some drawbacks, a quantitative measure that can quantify the complexity of EOPs is very necessary to compensate for them. In order to quantify the complexity of steps included in EOPs, Park et al. suggested the step complexity (SC) measure. In addition, to ascertain the appropriateness of the SC measure, averaged step performance time data obtained from emergency training records for the loss of coolant accident and the excess steam dump event were compared with estimated SC scores. Although averaged step performance time data show good correlation with estimated SC scores, conclusions for some important issues that have to be clarified to ensure the appropriateness of the SC measure were not properly drawn because of lack of backup data. In this paper, to clarify remaining issues, additional activities to verify the appropriateness of the SC measure are performed using averaged step performance time data obtained from emergency training records. The total number of available records is 36, and training scenarios are the steam generator tube rupture and the loss of all feedwater. The number of scenarios is 18 each. From these emergency training records, averaged step performance time data for 30 steps are retrieved. As the results, the SC measure shows statistically meaningful correlation with averaged step performance time data. In addition, since it is observed that the SC measure seems to have the procedure independent property (i.e. steps that have similar SC scores, whether they are included in different procedures or not, would have similar step performance time), it can be concluded that the SC measure can represent the complexity of steps included in EOPs.

In complex systems, such as nuclear power plants (NPPs) or airplane control systems, human errors play a major role in many accidents. Therefore, to prevent occurrences of accidents or to ensure system safety, extensive effort has been made to identify significant factors that cause human errors. According to related studies, written manuals or operating procedures are revealed as one of the most important factors, and complexity or understandability of a procedure is pointed out as one of the major reasons that make procedure-related human errors. Many qualitative checklists are suggested to evaluate emergency operating procedures (EOPs) of NPPs. However, since qualitative evaluations using checklists have some drawbacks, a quantitative measure that can quantify the complexity of EOPs is imperative to compensate for them. In order to quantify the complexity of EOPs, Park et al. suggested the step complexity (SC) measure to quantify the complexity of a step included in EOPs. In this paper, to ensure the appropriateness of the SC measure, SC scores are compared with averaged step performance time data obtained from emergency training records. The total number of available records is 36, and training scenarios are the loss of coolant accident and the excess steam dump event. The number of scenario is 18 each. From these emergency training records, step performance time data for 39 steps are retrieved, and they are compared with estimated SC scores of them. In addition, several questions that are needed to clarify the appropriateness of the SC measure are also discussed. As a result, it was observed that estimated SC scores and step performance time data have a statistically meaningful correlation. Thus, it can be concluded that the SC measure can quantify the complexity of steps included in EOPs.

For a nuclear power plant (NPP), symptom-based emergency operating procedures (EOPs) have been adopted to enhance the safety of NPPs through reduction of operators’ workload under emergency conditions. Symptom-based EOPs, however, could place a workload on operators because they have to not only identify related symptoms, but also understand the context of steps that should be carried out. Therefore, many qualitative checklists are suggested to ensure the appropriateness of steps included in EOPs. However, since these qualitative evaluations have some drawbacks, a quantitative measure that can roughly estimate the complexity of EOP steps is imperative to compensate for them. In this paper, a method to evaluate the complexity of an EOP step is developed based on entropy measures that have been used in software engineering. Based on these, step complexity (SC) measure that can evaluate SC from various viewpoints (such as the amount of information/operators’ actions included in each EOP step, and the logic structure of each EOP step) was developed. To verify the suitableness of the SC measure, estimated SC values are compared with subjective task load scores obtained from the NASA-TLX (task load index) method and step performance time obtained from a full scope simulator. From these comparisons, it was observed that estimated SC values generally agree with the NASA-TLX scores and step performance time data. Thus, it could be concluded that the developed SC measure would be considered for evaluating SC of an EOP step.

Emergency operating procedures (EOPs) in nuclear plants guide operators in handling significant process disturbances. Historically these procedures have been paper-based. More recently, computer-based procedure (CBP) systems have been developed to improve the usability of EOPs. The objective of this study was to establish human factors review guidance for CBP systems based on a technically valid methodology. First, a characterization of CBPs was developed for describing their key design features, including both procedure representation and functionality. Then, the research on CBPs and related areas was reviewed. This information provided the technical basis on which the guidelines were developed. For some aspects of CBPs the technical basis was insufficient to develop guidance; these aspects were identified as issues to be addressed in future research.

The current emergency operating procedures (EOPs) for a three-loop pressurized water reactor on steam generator tube rupture (SGTR) incident are experimentally evaluated at the Institute of Nuclear Energy Research Integral System Test Facility. The focus is on the adequacy of EOPs to limit primary-to-secondary leakage with particular emphasis on the number of ruptured U-tubes on the severity of an incident and the response time available for operator actions. By comparing test data with plant records of a Mihama-2 SGTR event, the key thermal-hydraulic phenomena during SGTR transients can be successfully simulated. The test results indicated that the current EOPs can function well in terminating the break flow and maintaining adequate core cooling. However, the effectiveness in minimizing the radioactive release demands an early and substantial operator involvement. To mitigate the consequences of such an event, the timing to isolate the faulted steam generator (SG) and to terminate safety injection flow will be critical. Furthermore, to avoid overfilling, the water level in the faulted SG needs to be drained prior to the implementation of the cooldown process.

Δημοσίευση σε επιστημονικό περιοδικό Summarization: As a result of their increasing complexity, modern industrial systems have come to rely on written procedures for dealing with various emergencies that might be encountered. Despite the significant effort invested in developing emergency operating procedures, traditional hard-copy procedures have been found inadequate in presenting complex instructions, handling cross-references, tracing suspended or incomplete steps, and in general, monitoring user progress. Developments in information technology have presented new opportunities for resolving these problems in using and adapting procedures. This article reviews current developments and problems with computer-based procedures and proposes a framework for specifying usability aspects of their use. The usability framework addresses critical cognitive activities involved in managing system emergencies and draws upon empirical studies of 'official' and 'selective' uses of procedures in actual practice. Finally, information technology applications are considered within the wider organizational context, such as compliance with procedures, crew communication, and retention of skills. Παρουσιάστηκε στο: Behaviour and Information Technology

New methods of information presentation and interface design have changed the work conditions in the modern nuclear power plant control room. One area that is receiving considerable attention is that of procedures—in particular, emergency operating procedures (EOPs). Such procedures are traditionally available as printed documents; however, developments in information technology have created new opportunities for procedure presentation, and this paper reports on the development of human factor guidelines for computerized presentation of EOPs. After identifying the principle stages in the transition from procedures as documents to fully automated procedures, computerized procedure presentation (CPP) is briefly discussed. Guidelines for the presentation of such are outlined, starting with the high level goals that also constitute the criteria against which the computerized procedures are measured during implementation. Six dimensions that describe CPP are presented and explored in detail, by identifying points in each dimension that characterize different levels of information technology sophistication.