No description available.

A series of small break loss of coolant accident (SBLOCA) analyses in nuclear plant shutdown operations was simulated using the code RELAP5A,MOD3 version 8.0 to predict the SBLOCA phenomena in the Zion-l nuclear power plant The first objective is to study the impact of SBLOCA (1" and 2" breaks) on plant conditions while in the shutdown mode. In particular, to determine the time to "core uncovery" without operator interaction. The other objective is to study the effect of RHR heat exchanger elevation on natural circulation mass flow rate, fluid temperature and peak fuel pin temperature. Peak temperature and time to core uncovery were found for two small break LOCA cases. The natural circulation mass flow rate after break initiation was affected by varying the RHR heat exchanger elevation. The system pressure and temperature were not affected much by the elevation change in the RHR heat exchanger. The current version of RELAP5/MOD3 was found to be sensitive to the initial conditions in studies of low pressure,low temperature plant systems, especially for a large break LOCA. / Graduation date: 1994

No description available.

Pressurizer modeling for predicting the dynamic pressure of the PIUS system is presented. The transient behavior of this model for the PIUS system was investigated. The validity of this model for the PIUS system is limited to transients that are neither too large nor too long in duration. For example, the model is not capable of describing events following a complete loss of liquid for the pressurizer. However, the model can be used for qualitative prediction of the PIUS system behavior for a wide variety of severe transients. A review of pressurizer modeling indicates that the neglecting of the change in the internal energy of the subcooled water during transients is an acceptable assumption. The inherently safe feature of the PIUS system was confirmed through the self-shutdown of the reactor or, in some cases, through reactor power reduction as a result of the ingress of the pool boric acid solution into the primary system. This dynamic model was constructed of three major components: (1) The primary loop, (2) The secondary loop, and (3) The natural convection loop through the pool. A lumped parameter model, uniform heat transfer, and point kinetics have been the main approximations in this model. Other approximations are mentioned during the modeling of each component of the model. The dynamic model was simulated using the DARE-P continuous system simulation language which was developed in the Electrical Engineering Department at the University of Arizona.

Typescript (photocopy). / Digitized by Kansas Correctional Industries

Thesis (Master of Engineering in Electrical Engineering)--Cape Peninsula University of Technology, 2018. / Worldwide a movement is underway to replace the burning of limited fossil fuel reserves for power generation with a cleaner, more efficient, yet still reliable and cost-effective method. Even though renewable technologies are often among the most common proposed, they are still limited by factors such as cost when considering large scale generation. Further requirements for replacing fossil fuel generation methods include the need to provide a continuous and reliable output for base load requirements, which is difficult to guarantee when making use of renewables alone. The proposed alternative is nuclear energy, as it is a reliable and cleaner method of power generation as compared to fossil fuels, capable of providing cost effective energy in the long run. The downside to nuclear energy, however, is the negative perception and general dislike of this method of generation, especially among the public who have been around this technology since its early days of implementation. The aim of this study is, therefore, to inform and prove that nuclear technology has evolved and come a long way since its early days, by making use of advanced technological solutions to address the fears associated with this technology from many years ago. The study further aims to prove that technologies such as advanced safety systems, new generations of reactors, advanced containment structures for both reactors and waste containment, as well as new waste disposal methods, have evolved nuclear energy into a safer and cleaner alternative method of power generation. This is achieved by first considering the origin of the negative perceptions surrounding the technology, and the nuclear accidents of the past, which have greatly influenced opinions about nuclear technology even up until today. After identifying the concerns and fears surrounding nuclear energy, research was conducted concerning how the latest technologies and innovations in safety systems are used to address these concerns, and ultimately eliminate the threats where possible. With the biggest concern identified, namely a core meltdown event leading to the release of radioactive material into the environment, two simulations were conducted to illustrate the unlikelihood of such an event occurring. The purpose of these simulations was, moreover, to illustrate the complexity and reliability of the various safety systems incorporated into the design of a nuclear power plant, preventing such a feared release of radioactivity from occurring. The research also importantly revealed that the dangers and possible threats posed by nuclear technology are often grossly overestimated, as under normal operating conditions a coal power plant, in fact, releases more radiation into the environment than a nuclear power plant. Further research reveals that the feared nuclear waste, produced by the nuclear industry yet regulated and disposed of properly, is only a small fraction of the highly hazardous waste produced on an annual basis worldwide. It is also revealed that in terms of fatalities, fossil fuel generation, on average, is responsible for more deaths annually than the biggest nuclear disasters that have ever occurred. Addressing the fears and concerns surrounding nuclear technology is therefore important, as this valuable resource may otherwise remain under-appreciated and under-utilised because of the misperceptions which currently exist amongst the public. This furthermore results in the unnecessary exhaustion of fossil fuel reserves, and concomitant pollution of the environment – all due to antiquated fears surrounding nuclear power plants.

Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1982. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Vita. / Bibliography: Ref 1-Ref 7. / by Walter Herbert Strohmayer. / Ph.D.

Prior to this work, MASSVAC (MASS eVACuation) had evolved as a micro-computer simulation model for analysis and evaluation of areas facing natural disasters (hurricanes and floods). Conceptual and technical enhancements have been made to procedures within MASSVAC to deal with the special problems of evacuating around nuclear power stations. Its incorporation into TEDSS-3 (Transportation-Evacuation Decision Support System) has resulted in a powerful tool to assist development of evacuation plans for nuclear power plants. The computer package comprehensively provides for all functions related to evacuation planning such as development of a socioeconomic and highway network database, estimation of evacuation time and development/evaluation of traffic management strategies to reduce network clearance times and to improve highway network performance during evacuation. Primary focus is on the new features incorporated into MASSVAC, especially in the trip distribution and traffic assignment procedures. Significant improvements have been made to the software implementations of the Dial traffic assignment and other key algorithms used in MASSVAC. The information content of the model's output has been enhanced for better understanding of the evacuation process and presentation of results. / Master of Science

Thesis (MBA)--Stellenbosch University, 2002. / ENGLISH ABSTRACT: The Pebble Bed Modular Reactor is a new and inherently safe concept for a nuclear power generation plant. In order to obtain the necessary licenses to build and operate this reactor, numerous design and safety analyses need to be performed. The results of these analyses must be supported with substantial proof to provide the nuclear authorities with a sufficient level of confidence in these results to be able to supply the required licences. Beside the obvious need for a sufficient level of confidence in the safety analyses, the analyses concerned with investment protection also need to be reliable from the investors’ point of view. The process to be followed to provide confidence in these analyses is the verification and validation process. It is aimed at presenting reliable material against which to compare the results from the simulations. This material for comparison will consist of a combination of results from experimental data, extracts from actual plant data, analytical solutions and independently developed solutions for the simulation of the event to be analysed. Besides comparison with these alternative sources of information, confidence in the results will also be built by providing validated statements on the accuracy of the results and the boundary conditions with which the simulations need to comply. Numerous standards exist that address the verification and validation of computer software, for instance by organisations such as the American Society of Mechanical Engineers (ASME) and the Institute of Electrical and Electronics Engineers (IEEE). The focal points of the verification and validation of the design and safety analyses performed on typical PBMR modes and states, and the requirements imposed by both the local and overseas nuclear regulators, are not entirely enveloped by these standards. For this reason, PBMR developed a systematic and disciplined approach for the preparation of the Verification and Validation Plan, aimed at capturing the essence of the analyses. This approach aims to make a definite division between software development and the development of technical analyses, while still using similar processes for the verification and validation. The reasoning behind this is that technical analyses are performed by engineers and scientists who should only be responsible for the verification and validation of the models and data they use, but not for the software they are dependent on. Software engineers should be concerned with the delivery of qualified software to be used in the technical analyses. The PBMR verification and validation process is applicable to both hand calculations and computer-aided analyses, addressing specific requirements in clearly defined stages of the software and Technical Analysis life cycle. The verification and validation effort of the Technical Analysis activity is divided into the verification and validation of models and data, the review of calculational tasks, and the verification and validation of software, with the applicable information to be validated, captured in registers or databases. The resulting processes are as simple as possible, concise and practical. Effective use of resources is ensured and internationally accepted standards have been incorporated, aiding in faith in the process by all stakeholders, including investors, nuclear regulators and the public. / AFRIKAASE OPSOMMING: Die Modulêre Korrelbedreaktor is ’n nuwe konsep vir ’n kernkragsentrale wat inherent veilig is. Dit word deur PBMR (Edms.) Bpk. ontwikkel. Om die nodige vergunnings om so ’n reaktor te kan bou en bedryf, te bekom, moet ’n aansienlike hoeveelheid ontwerp- en veiligheidsondersoeke gedoen word. Die resultate wat hierdie ondersoeke oplewer, moet deur onweerlegbare bewyse ondersteun word om vir die owerhede ’n voldoende vlak van vertroue in die resultate te gee, sodat hulle die nodigde vergunnings kan maak. Benewens die ooglopende noodsaak om ’n voldoende vlak van vertroue in die resultate van die veiligheidsondersoeke te hê, moet die ondersoeke wat met die beskerming van die beleggers se beleggings gepaard gaan, net so betroubaar wees. Die proses wat gevolg word om vertroue in die resultate van die ondersoeke op te bou, is die proses van verifikasie en validasie. Dié proses is daarop gerig om betroubare vergelykingsmateriaal vir simulasies voor te lê. Hierdie vergelykingsmateriaal vir die gebeurtenis wat ondersoek word, sal bestaan uit enige kombinasie van inligting wat in toetsopstellings bekom is, wat in bestaande installasies gemeet is, wat analities bereken is; asook dit wat deur ’n derde party onafhanklik van die oorspronklike ontwikkelaars bekom is. Vertroue in die resultate van die ondersoeke sal, behalwe deur vergelyking met hierdie alternatiewe bronne van inligting, ook opgebou word deur die resultate te voorsien van ’n gevalideerde verklaring wat die akkuraatheid van die resultate aantoon en wat die grensvoorwaardes waaraan die simulasies ook moet voldoen, opsom. Daar bestaan ’n aansienlike hoeveelheid internasionaal aanvaarde standaarde wat die verifikasie en validasie van rekenaarsagteware aanspreek. Die standaarde kom van instansies soos die Amerikaanse Vereniging vir Meganiese Ingenieurs (ASME) en die Instituut vir Elektriese en Elektroniese Ingenieurs (IEEE) – ook van Amerika. Die aandag wat deur die Suid-Afrikaanse en oorsese kernkragreguleerders vereis word vir die toestande wat spesifiek geld vir korrelbedreaktors, word egter nie geheel en al deur daardie standaarde aangespreek nie. Daarom het die PBMR maatskappy ’n stelselmatige benadering ontwikkel om verifikasie- en validasieplanne voor te berei wat die essensie van die ondersoeke kan ondervang. Hierdie benadering is daarop gemik om ’n duidelike onderskeid te maak tussen die ontwikkeling van sagteware en die ontwikkeling van tegniese ondersoeke, terwyl steeds gelyksoortige prosesse in die verifikasie en validasie gebruik sal word. Die rede hiervoor is dat tegniese ondersoeke uitgevoer word deur ingenieurs en wetenskaplikes wat net vir verifikasie en validasie van hulle eie modelle en die gegewens verantwoordelik gehou kan word, maar nie vir die verifikasie en validasie van die sagteware wat hulle gebruik nie. Ingenieurs wat spesialiseer in sagteware-ontwikkeling behoort verantwoordelik te wees vir die daarstelling van sagteware wat deur die reguleerders gekwalifiseer kan word, sodat dit in tegniese ondersoeke op veiligheidsgebied gebruik kan word. Die verifikasie- en validasieproses van die PBMR is sowel vir handberekeninge as vir rekenaarondersteunde-ondersoek geskik. Hierdie proses spreek spesifieke vereistes in onderskeie stadiums gedurende die lewenssiklusse van die ontwikkeling van sagteware en van tegniese ondersoeke aan. Die verifikasie- en validasiewerk vir tegniese ondersoeksaktiwiteite is verdeel in die verifikasie en validasie van modelle en gegewens, die nasien van berekeninge en die verifikasie en validasie van sagteware, waarby die betrokke inligting wat gevalideer moet word, versamel word in registers of databasisse. Die prosesse wat hieruit voortgevloei het, is so eenvoudig as moontlik, beknop en prakties gehou. Hierdeur is ’n effektiewe benutting van bronne verseker. Internasionaal aanvaarde standaarde is gebruik wat die vertroue in die proses deur alle betrokkenes, insluitende beleggers, die owerhede en die publiek, sal bevorder.

With the ever advancing state of computer systems, it is imperative to maintain the most up-to-date and reliable safety evaluation data for nuclear power systems. Commonplace now is the practice of updating old accident simulation results with more advanced models and codes using today's faster computer systems. Though it may be quite an undertaking, the benefits of using a more advanced model and code can be significant especially if the result of the new analysis provides increased safety margin for any plant component or system. A series of parametric and sensitivity studies for the Loss of Normal Feedwater Anticipated Transient without Scram (LONF ATWS) for Southern Company's Vogtle Electric Generating Plant (VEGP) Units 1&2 located near Waynesboro, GA was performed using the best-estimate thermal-hydraulics transient analysis code RETRAN-02w. This thesis includes comparison to the results of a generic plant study published by Westinghouse Electric Corporation in 1974 using an earlier code, LOFTRAN, as well as Vogtle-specific analysis. The comparative analysis exposes and seeks to explain differences between the two codes whereas the Vogtle analysis utilizes data from the Vogtle FSAR to generate plant-specific data. The purpose of this study is to validate and update the previous analysis and gather more information about the plant actions taken in response to a LONF ATWS. As a result, now there is a new and updated evaluation of the LONF ATWS for both a generic 4-loop Westinghouse plant and VEGP using a more advanced code. Beyond the reference case analysis, a series of sensitivity and parametric studies have been performed to show how well each type of plant is designed for handling an ATWS situation. These studies cover a wide range of operating conditions to demonstrate the dependability of the model. It was found that both the generic 4-loop Westinghouse PWR system and VEGP can successfully mitigate a LONF ATWS throughout the core's operating cycle.