The computational ability to accurately simulate boiling water reactor operation under the full range of standard steady-state operation, along with the capability to fully track the isotopic distribution of any fueled region in any location in the core has been developed. This new three-dimensional node-by-node capability can help designers track, for example, a full suite of minor and major actinides, fission products, and even light elements that result from depletion, decay, or transmutations. This isotopic tracking capability is not restricted to BWRs and can be employed in the modeling of PWRs, CANDUs, and other reactor types that can be modeled with the NESTLE code, the base core simulator employed in this research. To accurately simulate boiling water reactor operation, a major thermal-hydraulics upgrade was performed which involved the implementation of a drift-flux solution scheme to model steady-state boiling water flow. Sub-cooled boiling and bulk boiling are accurately modeled and a scheme for computing the correct flow distribution has been implemented. In addition, the incorporation of a nodal ORIGEN-based microscopic depletion solution has been included which allows for exceptional detail in tracking a large number of elements in every node of a core design, thus accounting for spectral dependencies such as moderator density effects, moderator temperature effects, fuel temperature effects, as well as controlled or uncontrolled conditions. The results of this study show the excellent fidelity of the two-phase solution for accurately predicting the boiling of water when compared to experimental results. Likewise, the isotopic inventory results show near-identical agreement with the well-established and validated ORIGEN-based SCALE/TRITON isotopic depletion sequence. The aim of these developments is to eventually produce a publicly available three-dimensional core simulator capable of assessing detailed isotopic inventories, a capability particularly valuable for the evaluation of recycling scenarios and actinide management in a variety of reactor types and fuel designs.

In order to improve the efficiency of dry used nuclear fuel (UNF) storage, reduced reactivity methods are being developed for various reactor types and operating conditions. Sub-criticality must be maintained in the storage configuration and conservative computer simulations are used as the primary basis for loading the storage casks. Methodologies are now being developed to reduce the amount of modeling and computation in order to make conservative assessments of how densely fuel can be packed. The SCALE/TRITON (Standardized Computer Analyses for Licensing Evaluation / Transport Rigor Implemented with Time-dependent Operation for Neutronic Depletion) code system has been used to simulate boiling water reactor (BWR) operating conditions in order to investigate nuclide densities in UNF and how the use of control rod blades affect nuclide densities found in UNF. Rodded and unrodded operating cases for a fuel assembly have been used as bounding cases and are used as reference solutions in a piecewise data approximation methodology (PDA method). A variety of control blade insertion patterns have been used with the PDA method and simulated in TRITON in order to observe trends in nuclide densities with varying control blade use. The PDA method is compared with TRITON simulated data in order to evaluate the validity and accuracy of the PDA method. The PDA method gives very accurate results for fissile nuclides but is insufficient in treating densities as a function of burnup for fission products and fertile nuclides. Predicting nuclide densities from temporally balanced control blade insertion and withdrawal patterns is also a strength of the PDA method. The PDA method, however, is not capable of properly accounting for neutron spectral shifts and the behavior in nuclide densities brought about by the spectral shift or nuclide density saturation. Observing the causes for the shortcomings in the PDA method, a more robust methodology can be developed.

This study aims to improve the modeling of partial SCRAM in the BISON plant model for the Forsmark 2 nuclear reactor after power uprate. Validation of the BISON model against tests performed from March to May in 2013 have shown that this is one of the areas in which there is room for improvement. After partial SCRAM is performed, the model underestimates the reactor power, recirculation flow and steam flow when compared to the measurement data. In BISON the partial SCRAM is modeled using a relative control rod effect vector (ASC vector). The aim is to replace the old values in this vector to improve the model. The new model was shown to give an improved result for the reactor power, recirculation flow and steam flow. The study gives recommendations on how to apply the new model and what values of the relative control rod effect vector that can be used in the future.

Thesis (M.S.)--University of Wisconsin -- La Crosse, 1980. / Includes bibliographical references (leaves 105-111).

This study explores the behavior of a Boiling Water Reactor (BWR) during a turbine trip scenario initiated by the abrupt closure of the turbine stop valve. The RELAP5/Mod3.4 code is employed to make calculations using the Laguna Verde Nuclear Power Plant input model provided by Innovative Software Systems Company. The event sequences and initial boundary conditions are sourced from the Boiling Water Reactor Turbine Trip 2 Benchmark created by NEA. Results are subsequently compared against the benchmark values. In order to gauge the risk of a turbine trip event leading to elevated power, which could in turn cause Critical Heat Flux (CHF)-related issues in cladding temperature, a best-estimate case is developed. Our findings indicate that the closure of the turbine stop valve (TSV) resulted in a collapse of the void fraction within the reactor core. Although the core power doubled the initial level, the negative feedback mechanism effectively suppressed the power pulse. Throughout the transient phase, the maximum cladding temperature stayed below the CHF threshold, a fact attributable to the fuel's conductivity and the rapid progression of the transient. We further analyzed three hypothetical scenarios to test the computational boundaries of the plant model. The third scenario, which combines conditions from the first two, produced elevated outcomes (6500MW core power, 598K cladding temperature, and 7900kPa dome pressure) as expected. Notably, while the CHF limit remained unbreached in this scenario, literature reviews suggest potential core meltdown risks in subsequent stages of this calculation. Our sensitivity analyses determined that variations in the gamma heating coefficient or the maximum time step of the calculations have little to no impact on core power or peak cladding temperature. Conversely, we noted a significant reduction, approximately 35\%, in the power peak, underscoring the high sensitivity of the parameters to the initial triggering of the SCRAM mechanism. Our results also recommend rapid and early actuation of the BPV as a measure to dampen the pressure wave, consequently decreasing both the power peak and peak cladding temperatures. / Thesis / Master of Applied Science (MASc) / This research investigates the response of the Laguna Verde Boiling Water Reactor to a turbine trip event using the RELAP5/Mod3.4 thermal-hydraulic analysis code. From reactor safety perspective a best-estimate case is evaluated, as well as three additional hypothetical scenarios. Findings are compared with the Boiling Water Reactor Turbine Trip II Benchmark results. Additionally, sensitivity analyses focusing on plant parameters such as shutdown rod behavior, gamma heating coefficient, turbine stop valve, and steam bypass valve characteristics conducted to determine their impact on the results. Insights from these analyses aim to enhance safety protocols and refine best practices in boiling water reactor management.

Fuel channels are used to keep the components of a nuclear fuel bundle in place. The fuel channel of a new nuclear fuel which is being developed at Westinghouse has a complex geometry which creates challenges in the manufacturing process. FE simulations were developed of a two-stage forming process of the fuel channel. Three different simulations were developed, a simulation of the pre-bending, a simplified simulation of bending, and a combined simulation of the two-stage process of pre-bending and bending. The simulations were done in Ansys Workbench. The simulations of the pre-bending could be validated against experimental results. The simulations of the bending showed significant differences in results compared to experiments. A few different sources of error were investigated due to the difference in results. / Höljerör omsluter kärnbränslet i en kärnreaktor och håller komponenterna i bränslepaketet på plats. Höljeröret av ett nytt kärnbränsle som utvecklas på Westinghouse har en komplex geometri som skapar utmaningar i tillverkningsprocessen. FE simuleringar av en tvåstegsprocess av bockningen av höljeröret utvecklades. Tre typer av simuleringar utvecklades, en simulering av förbockningen, en förenklad simulering av bockningen, och en simulering av den kombinerade förbockningen och bockning. Simuleringarna gjordes i Ansys Workbench. Förbockningssimuleringarna kunde valideras mot experiment. Bockningssimuleringen visade dock signifikanta skillnader i resultat gentemot experiment. Några olika felkällor undersöktes på grund av skillnaderna i resultat.

Thesis (M.S.)--University of Wisconsin -- La Crosse, 1983. / Includes bibliographical references (leaves 47-51).

Orientador: João Manoel Losada Moreira / Dissertação (mestrado) - Universidade Federal do ABC, Programa de Pós-Graduação em Energia, 2011

In the process of a boiling water reactor, high-levels of waste and radiation occur, where almost all the dose per person of the radiation in Forsmark are due to the radioactive iso-tope cobalt-60. The reason is that the stable isotope cobalt-59 is converted to the radioac-tive isotope cobalt-60 due to neutron irradiation in the reactor. Since 2012, unusually high levels of cobalt-60 have been observed at Forsmark 2 which occurs from the material Stel-lite and is a very common sealant in valves. The major disadvantage of the material Stellite in nuclear power is the high concentration of cobalt-59. When grinding alloy surfaces con-taining Stellite, cobalt-59 is released in the form of abrasive dust if the effectiveness of sub-sequent Stellite alloys is poor. The consequences lead to increased radiation levels, which implies major financial costs and a difficult work environment in, for example, mainte-nance work.Today, there is no mapping of valves containing Stellite, which may result in the decon-tamination of Stellite not being requested and missing when a maintenance action in the form of, for example, grinding is performed. The completed mapping of valves containing Stellite is thus the first that has been carried out within Forsmarks Kraftgrupp AB for the priority systems 313, 321, 331 and 415.In this work, valves containing Stellite have been mapped along main lines in systems that come into contact with reactor water without passing ion exchange filters. Furthermore, the effects of how the grinding of valves alloy surfaces in the seat / cone affects the feeding of cobalt-59 into the reactor and the effectiveness of subsequent decontamination of Stel-lite after grinding was investigated.The work has been divided into two main moments; Status analysis and Mapping, which in turn is divided into several sub-moments. The status analysis gathered the information re-quired to perform the mapping. With the gathered information from the status analysis, mapping was then carried out and valves were inventoried in the priority systems.A total of 45 valves containing Stellite were found whose water flow is likely to end up in the reactor without passing ion exchange filters. A total of 13 valves containing Stellite were found, which are not detected by the chemical departments measurement points and whose waterflow did not pass ion exchange filters before the reactor for systems 321 and 331.During a decontamination of Stellite in a valve, only alloy surfaces in the valves are con-trolled and cleaned, which results in dust from grinding remaining in the other surfaces of the valve as well as in the pipe ends when the valve has been assembled prior to commis-sioning. Of the 45 valves containing Stellite which have been inventoried, grinding in theseat/cone have occurred in eight of the valves, but only two of the valves have been de-contaminated since 2010-01-01. Since no decontamination of Stellite has occured six of eight times after grinding, and only alloy surfaces are checked as well as decontaminated, the effectiveness of subsequent decontamination of Stellite after grinding is very low.Based on the results of the work, a number of improvement proposals have been present-ed for continued work to reduce the feeding of cobalt 59 to the reactor water and eventu-ally reduce the radiation levels at Forsmark's nuclear power plant. / Vid processen i en kokvattenreaktor uppstår högaktivt avfall och höga strålningsnivåer, där nästan all persondos av strålning på Forsmark beror av den radioaktiva isotopen kobolt-60. Anledningen är att den stabila isotopen kobolt-59 omvandlas till den radioaktiva isotopen kobolt-60 vid neutronbestrålning i reaktorn. Man har sedan 2012 noterat ovanligt höga halter av kobolt-60 på Forsmark 2 vilket härrör till materialet Stellite, som är ett mycket vanligt tätningsmaterial i ventiler. Den stora nackdelen med Stellite i kärnkraftssamman-hang är den höga koncentrationen av kobolt-59. Vid slipning av legeringsytor innehållande Stellite, riskeras kobolt-59 frigöras i form av slipdamm om effektiviteten av efterföljande Stellitesaneringar är dålig. Konsekvenserna leder till ökade strålningsnivåer vilket innebär stora ekonomiska kostnader och en försvårad arbetsmiljö vid till exempel underhållsar-beten.Idag finns ingen kartläggning av ventiler innehållande Stellite, vilket kan resultera i att Stellitesaneringar inte begärs och uteblir då en underhållsåtgärd i form av till exempel slipning utförs. Den genomförda kartläggningen av ventiler innehållande Stellite är där-med den första som har utförts inom Forsmarks Kraftgrupp AB för de prioriterade syste-men 313, 321, 331 och 415.I detta arbete har ventiler innehållande Stellite kartlagts längs huvudledningar i system som kommer i kontakt med reaktorvatten utan att passera jonbytarfilter. Vidare har effekterna av hur slipning av ventilers legeringsytor i säte/kägla påverkar inmatningen av kobolt-59 och effektiviteten av efterföljande Stellitesaneringar undersökts.Arbetet har delats upp i två huvudmoment; Nulägesanalys och Kartläggning, som i sin tur delats upp i flera delmoment. I nulägesanalysen samlades den information som krävdes för att utföra kartläggningen. Med den inhämtade informationen från nulägesanalysen, inven-terades och kartlades sedan ventiler i de prioriterade systemen.Totalt hittades 45 stycken ventiler innehållande Stellite vars vattenflöde riskerar att hamna i reaktorn utan att passera jonbytarfilter. Sammanlagt hittades 13 stycken ventiler innehål-lande Stellite som ej registreras av kemiavdelningens provtagningar och som inte passerar jonbytarfilter innan reaktorn för system 321 och 331.Vid en Stellitesanering kontrolleras och saneras endast legeringsytor i ventiler, vilket re-sulterar i att slipdamm kan finns kvar i ventilens övriga ytor samt i rörändarna då ventilen har monterats ihop inför driftsättning. Av de 45 stycken ventiler innehållande Stellite som har inventerats, har åtta stycken slipats i säte/kägla men enbart två stycken Stellitesanerats efter slipning sedan 2010-01-01. Eftersom Stellitesaneringar efter slipning har uteblivitsex av åtta gånger och endast legeringsytor kontrolleras samt Stellitesaneras, är effektivite-ten av efterföljande Stellitesaneringar vid slipning mycket låg.Baserat på resultaten av arbetet, har ett antal förbättringsförslag presenterats för fortsatt arbete att minska kobolt-59-inmatningen till reaktorvattnet och på sikt minska strål-ningsnivåerna på Forsmarks kärnkraftverk.

Examensarbetet behandlar problematik som uppstår vid härdsmältningsförlopp i en kärnkraftsreaktor av typen kokvattensreaktor. Resultatet ska användas som riktlinjer till strategier som utvecklas av Severe Accident Management Guidelines (SAMG) där arbetets uppdrag är ett delmoment i framtagning av strategier för att bemästra de olika fenomen som uppstår vid härdsmälta.   Syftet med arbetet är att undersöka begränsningar för att bevara reaktortanken intakt vid haveri, genom att undersöka den minsta mängd vatten som behövs för att undvika tankgenomsmältning. Undersöka fallen som leder till härdsmälta och gruppera dem efter händelsesekvenser. Undersöka metall/vatten-reaktionen som uppstår då härden blir över 800°C och undersök om fallen kan grupperas i händelsesekvenser.  Metoder som används i arbetet är PSA-dokumentation, händelseutvecklingsträd, teoretiska beräkningar och MAAPv5.03. Resultatet beskriver att grupperingar av fallen som slutar i härdsmälta och grupperingar av metall/vatten-reaktionen hos de olika fallen kan genomföras. Resultatet beskriver även ett minsta flöde som kan föras in i reaktortanken för att hindra tankgenomsmältning och flöden upp till 100 kg/s så det finns resultat för olika flöden beroende på vilka kylmedel som är tillgängliga.  Slutsatsen av arbetet är att fall kan grupperas efter händelsesekvenser och påverkan hos metall/vattenreaktion, grupperingarna sparar tid vid ett haveriförlopp. I varje grupp kunde det svåraste fallet beräknas för minsta flöde för att klara tankgenomsmältning och flöden upp till 100 kg/s.  Framtida arbeten bör undersöka trycket och vätgasen som skapas vid vatteninmatning samt dess påverkan på reaktorinneslutningen. / The master thesis deals with problems that arise during nuclear meltdown in a nuclear powerplant of the type boiling water reactor. The work will be used as guidelines for strategies developed by Severe Accident Management Guidelines (SAMG), this master thesis is a sub-element in the development of strategies for mastering the various phenomena that arise during a meltdown.  The purpose of the work is to investigate limitations for maintaining the reactor tank intact during the meltdown by, examining the minimum amount of water needed to avoid the meltdown getting through the reactor tank. Examining the cases that lead to meltdown and group them according to the event sequences. Examine the metal/water-reaction that occurs when the core becomes over 800°C and examine if the cases can be grouped into event sequences.  Methods used in the master thesis is PSA-analysis, event development threes, theoretical calculations and MAAPv5.03.  The result describes the groupings of the cases ending in meltdown and the groupings of the metal/water-reaction of the various cases. The result also describes a minimum flow that is required to prevent meltdown of getting through the reactor tank and flow up to a 100 kg/s.  The conclusion of the master thesis is that cases can be grouped according to event sequences and the influence of the metal/water-reaction, the groupings save time in the event of a breakdown. In each group the most difficult case was calculated so that the lowest flow to prevent the meltdown from getting through the reactor tank was presented among with different flows up to 100 kg/s. Future work should investigate the pressure and hydrogen gas created by the water input and its influence on the reactor inclusion.