THE CASE ANALYSIS OF UPPER BIG BRANCH MINE

This article aims to analyze a catastrophic mining explosion which resulted in 29 deaths in West Virginia, U.S. The first reports upon the explosion suggested that the explosion happened due to lack of appropriate safety measures. However, further investigation revealed that the issue was deeper than merely the absence of appropriate safety measures. The negative organizational culture created by the leadership was considered as the root cause of this catastrophic incident. According to a case study published by The U.S. Nuclear Regulatory Commission in 2012, it appeared that the organization made systematic and aggressive efforts to prioritize production over the safety of its employees. The disaster could have been prevented if the leadership had taken appropriate safety measures. Leadership, who can see the big picture, understands that prioritizing safety results in overall performance improvement in the long term (Krause, 2005). It is possible to see the implications of such leadership mindset in the organization’s culture. Showing workers that the organization will always do the right thing to assure their safety is an important step toward building trust across the board. Otherwise, lack of trust and communication may eventually lead to tragic incidents as in the case of the Massey Energy. The despotic leadership, that constantly imposed fear on its employees to discourage them from voicing their opinions and questioning the existing conditions, eventually brought organizational deviance. Members of the organization neither had any meaningful communication nor appropriate information exchange. The absence of mutual trust and respect in the work environment was apparent. This paper offers further insights into the role of leadership in the prevention of future catastrophic incidence while promoting both safety and enhanced performance. KEYWORDS: Inclusive, leadership, organization, production, safety

Characteristics of the Vertical and Horizontal Response Spectra of Earthquakes in the Jeju Island Region

In this study, we evaluated the response spectra of 24 earthquake series, which includes 15 from the Kumamoto earthquake series and 9 from the Pohang earthquake series, and explored the effects of earthquake magnitude on the resonance frequencies of structures and buildings. Furthermore, the observations of this study were compared with the design response spectra, Regulatory Guide 1.60 (The United States Nuclear Regulatory Commission, 1973) for Korean nuclear power plants, and with the Korean Building Code (MOLIT, 2016, hereinafter referred to as KBC 2016) for general structures and buildings. The response spectra, after normalization with reference to the peak ground acceleration (PGA), were derived using a total of 423 horizontal and vertical accelerations. It was observed that the shapes of the horizontal and vertical response spectra were strongly dependent on the magnitude of the earthquake and the resonance frequency. Given the strong dependence of the response on the magnitude, it is suggested to consider magnitude > ML ~ 6.0 when establishing design response spectra. Compared to inland areas, a fairly higher amplitude of response at significantly lower frequency ranges could be attributed to the local geological environment of Jeju Island, which was formed by a surface volcano eruption and the distribution of unconsolidated Pleistocene marine sediments in the Jeju area. It is necessary to study the characteristic influence of layers with low shear wave velocity distributed in the Jeju region on seismic responses more rigorously while considering the frequency band and amplitudes at the surface of Jeju. The resonance frequencies of general low-rise and mid-rise buildings by the brief formula and those by design response spectra both suggested by KBC 2016 were overlapped, and these indicated that the seismic hazard could be much higher on general buildings in the Jeju region than in inland areas. Lastly, it is necessary to make the design standard criteria for Reg. Guide 1.60 and KBC 2016 more conservative in the lower frequency range of higher than 0.6 Hz and 2.0–6.0 Hz, respectively, which is significantly lower than those of the inland area, and to establish improved design response spectra with site-specific seismic design standards by referencing large amounts of qualitative data from the region around the Korean Peninsula.

Controls on Exports of Nuclear Commodities and on Assistance to Foreign Nuclear Activities

Chapter 4 covers two related sets of U.S. government controls on nuclear-related items that flow from the Atomic Energy Act of 1954 and the Nuclear Non-Proliferation Act of 1978. One, administered by the Nuclear Regulatory Commission (NRC), covers exports of nuclear hardware and nuclear materials. The other, called “Part 810” and administered by the National Nuclear Security Administration (NNSA) of the U.S. Department of Energy, covers assistance by U.S. persons (including transfers of nuclear-related technology) to foreign nuclear activities. The chapter explains: which items and activities are subject to the NRC and NNSA regulations; the basis and criteria for their restrictions; how to determine whether your commodity or activity is covered and, if so, whether you will need a license to export or reexport it; how to get a license if one is required; and the potential penalties for violating the rules. The chapter also explains how the NRC and NNSA rules relate to the regulatory regimes covered in other parts of the book.

The Oriented Spray Cooling System for Heat Rejection and Evaporation

Spray ponds offer significant advantages over mechanical draft cooling towers including superior simplicity and operability, lower preferred power requirements, and lower costs. Unlike a conventional spray pond in which spray nozzles are arranged in a flat bed and water is sprayed upward, the Oriented Spray Cooling System (OSCS) is an evolutionary spray pond design in which nozzles are mounted on spray trees arranged in a circle and are tilted at an angle oriented towards the center of the circle. Therefore, each nozzle is exposed to essentially ambient air as water droplets drag air into the spray region while the warm air concentrated in the center of the circle rises. Both of these effects work together to increase air flow through the spray region. Increased air flow reduces the local wet-bulb temperature of the air in the spray pattern, promoting heat transfer and more efficient cooling. The authors have developed analytical models to predict the thermal performance of the OSCS that are based on classical heat and mass transfer and kinetic vector relationships for spherical water droplets that rely only on generic experimental thermal performance data. The model is not limited in application with regard to spray pressure or nozzle spacing or orientation and is not limited by droplet size considerations. The paper compares the predicted performance of the OSCS with full-scale field test results that were measured in compliance with Nuclear Regulatory Commission requirements at the Columbia Generating Station where the ultimate heat sink is two OSCS.

Human performance and system safety: Alternative perspectives on safety assessment

Discussion Panel Abstract: The recent Boeing 737MAX accidents crystalized for the public the complexity of anticipating system and operator performance and developing a system design that prevents catastrophic outcomes. The operational situations, progression of flightcrew actions, and system behaviors that led to the two accidents had not been anticipated by the manufacturer or the regulator. These accidents were only the most recent examples of our failure to anticipate and manage operational complexities and operator performance. The art and science of human factors has yet to perfect risk assessment (or safety assessment) for complex systems. In the not-so-distant past, system risk assessment made estimates of human error probabilities (HEPs) for specific operational tasks, which were combined with estimated equipment failure rates to produce an overall risk estimate. Indeed, these Human Reliability Analysis (HRA) techniques have evolved over decades and are still being developed (e.g., IDHEAS-ECA, Xing et al., 2020), partly because they satisfy the need for a simple quantitative threshold that can be used by industry and regulators: if risk probability is too high, change the design or some other aspect of operations. Through the years, there have been critiques of the HRA approach (e.g., Hollnagel, 1998) that led to revisions, such as focusing on cognitive functions instead of operator tasks, but not to the basic quantitative risk-estimation approach. Other approaches to assessing risk/safety have wandered down other paths: attempting to capture system complexity from an operator’s perspective (Roth, Mumaw, Lewis, 1994; Nuclear Regulatory Commission, 2000), or better documenting the many ways in which system operators manage complexity daily to find ways to improve their capacity (Hollnagel, Woods, & Leveson, 2006). These approaches have used different measures than HEPs; e.g., measures of operator performance, measures of interface usability/design, measures of task complexity, and the analysis of system constraints. In this panel, we offer different perspectives on risk/safety assessment as it relates to operator performance in complex systems. Foundational to assessment is deciding the nature of safety and the role of operator performance. Another important question is, as you move away from simple quantitative measures, how do you establish safety thresholds? That is, what guidance can we give to industry and regulators regarding how to measure safety and how to decide that action is required on the basis of safety.

Putting Gonuke Into Practice: Considerations for Human Factors Evaluations

The Guideline for Operational Nuclear Usability and Knowledge Elicitation (GONUKE) framework was introduced in 2015 to support human factors evaluations needed for control room upgrades at nuclear power plants. NUREG-0711, the Human Factors Engineering Program Review Model, is used by the U.S. Nuclear Regulatory Commission to review human factors activities associated with human-system interfaces at nuclear power plants, and GONUKE is anchored to the phases of development and design in NUREG-0711. This paper addresses five considerations to help users of GONUKE better apply the framework to evaluations for NUREG-0711 and beyond. These five considerations are: (1) GONUKE only specifies evaluation, not design; (2) GONUKE is a framework, not a method or process; (3) GONUKE goes beyond NUREG-0711 requirements; (4) GONUKE application shouldfollow a graded approach; (5) different evaluations are required fo r formative vs. summative phases.

Results of a Virtual Round Robin Study to Estimate Probability of Detection for Dissimilar Metal Welds

This paper presents efforts to overcome challenges with empirical probability of detection (POD) estimations in the nuclear power industry through the utilization of a novel virtual flaw method. A virtual round robin (VRR) study was conducted under the Program for Investigation Of NDE by International Collaboration (PIONIC), organized by the United States Nuclear Regulatory Commission (NRC) utilizing data generated by the virtual flaw method. Analysis of results from the VRR was performed by teams from Pacific Northwest National Laboratory (PNNL), Electric Power Research Institute (EPRI), and Aalto University. Empirically derived POD estimations are presented, and challenges associated with obtaining these estimations are discussed. The virtual flaw method is introduced and some details of its implementation for the VRR activity are described. Results from POD analysis of the VRR data by PNNL, EPRI, and Aalto University are presented and a discussion regarding differences in analysis results is provided. Finally, potential future efforts to improve the application of the virtual flaw method and its estimation of POD are discussed.

Quality assurance program for Angra 1 licence renewal and long-term operation

In Brazil, according to Cnen standard, a nuclear power plant has authorization to operate for 40 years. Angra 1 commercial operation started in 1985 and it has license to operate until 2024. Eletronuclear aims to extend the operation of the Angra 1 plant from 40 to 60 years. To obtain the license renewal by more than 20 years (long-term operation), Eletronuclear will need to meet the requirements of 10 CFR Part 54, Cnen NT-CGRC-007/18 and NT-CGRC-008/18 (Cnen technical notes). To obtain a license renewal to a long-term operation it is necessary to demonstrate that the plants will operate according to safety requirements, through analysis, testing, aging management, system upgrades, as well as additional inspections. Plant operators and regulators must always ensure that plant safety is maintained and, when it is possible, strengthened during the long-term operation of the plant. One of the documents to obtain a license renewal to a long-term operation is the Quality Assurance Program (QAP). Angra 1 has a QAP according to 10CFR 50 App B and Cnen NN 1.16 for safety related items. However, according to 10 CFR50.34, Nureg-1800 Appendix A.2, Nureg-1801 Appendix A-1 of Nuclear Regulatory Commission (NRC) and NT-CGRC-007/18 and NT-CGRC-008/18 of Cnen, the QAP needs to include the items that are not safety related but are included in the Aging Management. This article will discuss the Angra 1 QAP for the license renewal to a long-term operation according the standards approved by Cnen.

Probabilistic Fracture Mechanics Evaluation of a BWR Feedwater Nozzle

A probabilistic fracture mechanics (PFM) evaluation was performed for the nozzle blend radius and nozzle-to-shell weld of a boiling water reactor (BWR) feedwater nozzle using the PFM methodology in Electric Power Research Institute (EPRI) Boiling Water Reactor Vessel and Internals Program (BWRVIP) BWRVIP-108-A and BWRVIP-241-A, which are the technical basis for inspection relief in ASME Code Case N-702. Using a finite element model of the feedwater nozzle, stress analysis was performed for plant-specific piping loads and bounding transients, which were grouped by severity and projected cycle counts. Monte Carlo simulations were performed using the VIPER-NOZ (Vessel Inspection Program Evaluation for Reliability, including Nozzle) PFM software to determine probabilities of failure for the reactor pressure vessel (RPV) with an inspection population of 25% of the feedwater nozzles every ten years for sixty years of plant operation. The results show that the probabilities of failure for normal operation and low temperature over pressure (LTOP) event meet the acceptance criteria for RPV failure in NUREG-1806 by the U.S. Nuclear Regulatory Commission (NRC). Thus, there is potential to seek regulatory relief to reduce the inspection population of BWR feedwater nozzles from 100% to 25% every ten years using the technical basis of ASME Code Case N-702.