scholarly journals Digitizing Native American Collections

2017 ◽  
Vol 45 (1) ◽  
pp. 11
Author(s):  
Nora Stewart

Note from the editor: DttP has been featuring student papers for a number of years at this point—I remember talking about it in my documents class in library school. (For the record, I did not get nominated, which was probably the right choice, even if I did find my paper about the Nuclear Regulatory Commission fascinating.) But as technology changes the work of libraries, so has it changed the products that our students are producing. What follows is an example of a libguide created by a Nora Stewart, student from Emporia State University. Although my presentation of this work is somewhat clumsy, I highly recommend that all of you visit the page (http://libguides.emporia.edu/c.php?g=531631&p=3637209) and look through the resources Nora has collected and placed into context.

2013 ◽  
Vol 2013 ◽  
pp. 1-12
Author(s):  
Maria Avramova ◽  
Diana Cuervo

Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.


Author(s):  
John P. McCloskey ◽  
Richard J. Smith

One of the requirements for validating nuclear reactor plant models is to benchmark the predicted results of selected transients against measured plant data or another qualified code. A major challenge with benchmarking is the criteria for validating a model are not always well defined and rely heavily on human judgment, thus introducing the possibility of human bias or inconsistent conclusions. The validation process can also be time consuming. A new method is presented to aid in the validation of nuclear reactor plant models, using the Automated Code Assessment Program (ACAP), which is a tool developed at Pennsylvania State University under contract by the U. S. Nuclear Regulatory Commission (NRC). The proposed method was developed specifically for real-time best-estimate nuclear operator training simulator transients. However, the tool can be easily customized for most applications (e.g., design models, steady state data). Four distinct statistical metrics and weightings were chosen, as deemed appropriate for transient nuclear operator training simulator applications. The metrics account for errors in magnitude and trend, and incorporate an experimental uncertainty. The four metrics are then combined into a single Figure of Merit (i.e., a statistical level of agreement between the predicted and benchmarking data sets). The use of ACAP for benchmarking is demonstrated by comparing experimental data from the Loss-of-Fluid-Test (LOFT) facility Large Break Loss-of-Coolant Experiment L2-5 with code predictions from a RELAP5-3D (Version 2.9.3+) model previously developed and published by Idaho National Laboratories. The proposed method is shown to have several advantages over conventional validation methods, in that it greatly reduces the possibility of human bias, generates reproducible results, can be easily automated to improve efficiency, and can be easily documented. After the initial validation, the tool can also be used to re-validate models after computer hardware changes, model changes, tool upgrades, and operating system upgrades.


Author(s):  
Mira Kabze

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


Author(s):  
Ryann E. Rupp ◽  
Michael D. McMurtrey

Abstract The Nuclear Regulatory Commission (NRC) is currently reviewing Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for potential endorsement. A prior NRC-sponsored review on the application of Section III, Division 5 design rules for elevated-temperature reactors have identified issues currently not in the code that need to be addressed. These include an insufficient knowledge of the effect of notches, structural discontinuities, and multiaxial stress. This work qualitatively evaluates Alloy 617 as notch strengthening or notch weakening and investigates the impact of multiaxial stress on Alloy 617 creep behavior. Multiaxial stress does not degrade Alloy 617’s short-term creep-rupture properties. This was concluded from the following observations: 1) multiaxial creep resided on or to the right of the Larson-Miller curve for uniaxial creep-rupture data, 2) Alloy 617 exhibited notch strengthening, and 3) creep-rupture life increased with a stronger multiaxial stress. Long-term and intermediate creep-rupture tests are in progress in order to determine whether notch behavior will cross from strengthening to weakening at lower stresses and longer rupture lifetimes. A nondestructive characterization methodology using X-ray computed tomography is explored as a means to identify the failure location on the specimen prior to rupture.


Author(s):  
Shawn Comstock ◽  
Mark Gowin

Several new reactors are currently under construction in the USA. Based on current construction schedules, Watts Bar 2 will be the first new reactor to go online for commercial generation since Watts Bar 1 was issued its operating license in 1996. New engineering programs will be going online with new reactors like Watts Bar 2. The startup of these new engineering programs is not without its own set of challenges. One of the programs has undergone a significant transformation since the last nuclear power plant started commercial operation in terms of industry implementation methods and regulatory requirements. In 1996, the NRC issued Generic Letter 96-05 to communicate issues related to periodic verification (PV) of motor-operated valves (MOVs) and to request action by operating commercial power reactors to establish an MOV PV program. Subsequently, the regulations were revised to include a requirement to have an MOV PV program in Title 10, “Energy,” of the Code of Federal Regulations (10 CFR) 50.55a(b)(3)(ii). Generic Letters 89-10 (on MOV surveillance and testing) and 96-05 have been closed and today stand as historical references. Their provisions do not directly apply to new reactors, but there are many lessons available from MOV PV programs at operating sites in terms of safety, implementation, and cost. There is only one consensus standard available to describe the requirements for an acceptable MOV PV program. This is contained in the ASME’s Operation and Maintenance of Nuclear Power Plants (OM Code) as Mandatory Appendix III. The U.S. Nuclear Regulatory Commission (NRC) previously endorsed this approach as a Code Case and is preparing a proposed change to 10 CFR 50.55a to incorporate by reference the ASME OM Code edition that includes Appendix III. This paper conveys the technical complexities and financial concerns faced by plant staff in making the right technical decisions for new program implementation at a new reactor in the USA. Paper published with permission.


2020 ◽  
Vol 1 (1) ◽  
pp. 11-18
Author(s):  
M. A. Rodionov ◽  
I. V. Akimova

In the submitted study the problem of the formation of financial literacy of students at informatics lessons and relevant training of future informatics teachers is considered. Financial literacy is understood as a set of basic knowledge in the field of finance, banking, insurance, as well as budgeting for personal finances that allow a person to choose the right financial product or service, soberly assess and take risks that may arise during the use of these products, correctly accumulate savings and identify doubtful (fraudulent) investment schemes. The authors conclude that successful development of meaningful lines of the course of financial literacy requires integration of a few school subjects, such as mathematics, history, informatics, social science and literature. The role of modern informatics teacher in the formation of financial literacy of students is great. Therefore, in the training of a future informatics teacher, it should be paid the attention to issues related to the study of elements of financial literacy in informatics lessons. In order to solve the problem, the authors propose to use the special course “Basics of work in 1С:Enterprise”, which is implemented at Penza State University. The article contains a program of the course and the methodological recommendations for its implementation.


2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.


Author(s):  
John O’Hara ◽  
Stephen Fleger

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.


Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton ◽  
Harold Adkins ◽  
Judith Cuta ◽  
Nicholas Klymyshyn ◽  
...  

In 2007, a severe transportation accident occurred near Oakland, California, at the interchange known as the “MacArthur Maze.” The accident involved a double tanker truck of gasoline overturning and bursting into flames. The subsequent fire reduced the strength of the supporting steel structure of an overhead interstate roadway causing the collapse of portions of that overpass onto the lower roadway in less than 20 minutes. The US Nuclear Regulatory Commission has analyzed what might have happened had a spent nuclear fuel transportation package been involved in this accident, to determine if there are any potential regulatory implications of this accident to the safe transport of spent nuclear fuel in the United States. This paper provides a summary of this effort, presents preliminary results and conclusions, and discusses future work related to the NRC’s analysis of the consequences of this type of severe accident.


Author(s):  
J. Xu ◽  
C. Miller ◽  
C. Hofmayer ◽  
H. Graves

Motivated by many design considerations, several conceptual designs for advanced reactors have proposed that the entire reactor building and a significant portion of the steam generator building will be either partially or completely embedded below grade. For the analysis of seismic events, the soil-structure interaction (SSI) effect and passive earth pressure for these types of deeply embedded structures will have a significant influence on the predicted seismic response. Sponsored by the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) is carrying out a research program to assess the significance of these proposed design features for advanced reactors, and to evaluate the existing analytical methods to determine their applicability and adequacy in capturing the seismic behavior of the proposed designs. This paper summarizes a literature review performed by BNL to determine the state of knowledge and practice for seismic analyses of deeply embedded and/or buried (DEB) nuclear containment type structures. Included in the paper is BNL’s review of the open literature of existing standards, tests, and practices that have been used in the design and analysis of DEB structures. The paper also provides BNL’s evaluation of available codes and guidelines with respect to seismic design practice of DEB structures. Based on BNL’s review, a discussion is provided to highlight the applicability of the existing technologies for seismic analyses of DEB structures and to identify gaps that may exist in knowledge and potential issues that may require better understanding and further research.


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