Дисертації з теми "BWR1"
Оформте джерело за APA, MLA, Chicago, Harvard та іншими стилями
Ознайомтеся з топ-50 дисертацій для дослідження на тему "BWR1".
Біля кожної праці в переліку літератури доступна кнопка «Додати до бібліографії». Скористайтеся нею – і ми автоматично оформимо бібліографічне посилання на обрану працю в потрібному вам стилі цитування: APA, MLA, «Гарвард», «Чикаго», «Ванкувер» тощо.
Також ви можете завантажити повний текст наукової публікації у форматі «.pdf» та прочитати онлайн анотацію до роботи, якщо відповідні параметри наявні в метаданих.
Переглядайте дисертації для різних дисциплін та оформлюйте правильно вашу бібліографію.
Raub, Sebastian. "Transient behaviour in a BWR with Hafnium Cladding : Feasibility study of using BWRs as Higher Actinide Burners at the Example of Ringhals I." Thesis, KTH, Skolan för teknikvetenskap (SCI), 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-38189.
Повний текст джерелаChun, John Hwan. "Modeling of BWR water chemistry." Thesis, Massachusetts Institute of Technology, 1990. http://hdl.handle.net/1721.1/13660.
Повний текст джерелаSoma, Kovács István. "Simplified Simulator for BWR Instabilities." Thesis, KTH, Fysik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-210626.
Повний текст джерелаFerroni, Paolo Ph D. Massachusetts Institute of Technology. "Steady state thermal hydraulic analysis of hydride fueled BWRs." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41263.
Повний текст джерелаThesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006.
(cont.) Since the results obtained in the main body of the analysis account only for thermal-hydraulic constraints, an estimate of the power reduction due to the application of neutronic constraints is also performed. This investigation, focused only on the "New Core" cases, is coupled with an increase of the thickness of the gap separating adjacent bundles from 2 to 5 mm. Under these more conservative conditions, the power gain percentages are lower, ranging between 24% and 43% (depending on the discharge burnup considered acceptable) for the upper pressure drop limit, and between 17% and 32% for the lower pressure drop limit.
(cont.) The benefits of the latter approach are evident since the space occupied by the bypass channel for cruciform control rod insertion becomes available for new fuel and a higher power can be achieved. The core power is constrained by applying thermal-hydraulic limits that, if exceeded, may induce failure mechanisms. These limits concern Minimum Critical Power Ratio (MCPR), core pressure drop, fuel average and centerline temperature, cladding outer temperature and flow-induced vibrations. To limit thermal-hydraulic instability phenomena, core power and coolant flow are constrained by fixing their ratio to a constant value. In particular, each BWR/5 core has been analyzed twice, each time with a different pressure drop limit: a lower limit corresponding to the pressure drop of the reference core and an upper limit 50% larger. It has been demonstrated that, in absence of neutronic constraints and with the maximum allowed pressure drop fixed at the upper limit, the implementation of the hydride fuel yields power gain percentages, with respect to oxide cores chosen as reference, of the order of 23% when its implementation is performed following the "Backfit" approach and even higher (50-70%) when greater design freedom is allowed in the core design, i.e. in the "New Core" approach. Should the maximum allowed pressure drop be fixed at the lower limit, the power gain percentage of the "Backfit" approach would decrease to 17%, while that of the "New Core" approach would remain unchanged, i.e. 50-70%.
This thesis contributes to the Hydride Fuel Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in Light Water Reactors (LWRs). Considerable work has already been accomplished on hydride fueled Pressurized Water Reactor (PWR) cores. This thesis extends the techniques used in the PWR analysis to examine the potential power benefits resulting from the implementation of the hydride fuel in Boiling Water Reactors (BWRs). This work is the first step towards the achievement of a complete understanding of the economic implications that may derive from the use of this new fuel in BWR applications. It is a whole core steady-state analysis aimed at comparing the power performance of hydride fueled BWR cores with those of typical oxide-fueled cores, when only thermal-hydraulic constraints are applied. The integration of these results with those deriving from a transient analysis and separate neutronic and fuel performance studies will provide the data required to build a complete economic model, able to identify geometries offering the lowest cost of electricity and thus to provide a fair basis for comparing the performance of hydride and oxide fuels. Core design is accomplished for two types of reactors: one smaller, a BWR/5, which is representative of existing reactors, and one larger, the ESBWR, which represents the future generation of BWRs. For both, the core design is accomplished in two ways: a "Backfit" approach, in which the ex-bundle core structure is identical to that of the two reference oxide cores, and a "New Core" approach, in which the control rods are inserted into the bundles in the form of control fingers and the gap between adjacent bundles is fixed optimistically at 2 mm.
by Paolo Ferroni.
S.M.
Morra, Paolo. "Design of annular fuel for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/34448.
Повний текст джерелаIncludes bibliographical references (p. 94).
Enabling high power density in the core of Boiling Water Reactors (BWRs) is economically profitable for existing or new reactors. In this work, we examine the potential for increasing the power density in BWR plants by switching from the current solid fuel to annular fuel cooled both on its inside and outside surfaces. The GE 8x8 bundle dimensions and fuel to moderator ratio are preserved as a reference to enable applications in existing reactors. A methodology is developed and VIPRE code calculations are performed to select the best annular fuel bundle design on the basis of its Critical Power Ratio (CPR) performance. Within the limits applied to the reference solid fuel, the CPR margin in the 5x5 and 6x6 annular fuel bundles is traded for an increase in power density. It is found that the power density increase with annular fuel in BWRs may be limited to 23%. This is smaller than possible for PWRs due to the different mechanisms that control the critical thermal conditions of the two reactors. The annular fuel could still be a profitable alternative to the solid fuel due to neutronic and thermal advantages.
by Paolo Morra.
S.M.
Karahan, Aydin. "An evolutionary fuel assembly design for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41304.
Повний текст джерелаIncludes bibliographical references (p. 138-140).
An evolutionary BWR fuel assembly design was studied as a means to increase the power density of current and future BWR cores. The new assembly concept is based on replacing four traditional assemblies and large water gap regions with a single large assembly. The traditional BWR cylindrical UO2-fuelled Zr-clad fuel pin design is retained, but the pins are arranged on a 22x22 square lattice. There are 384 fuel pins with 9.6 mm diameter within a large assembly. Twenty-five water rods with 27 mm diameter maintain the moderating power and accommodate as many finger-type control rods. The total number and positions of the control rod drive mechanisms are not changed, so existing BWRs can be retrofitted with the new fuel assembly. The technical characteristics of the large fuel assembly were evaluated through a systematic comparison with a traditional 9x9 fuel assembly. The pressure, inlet subcooling and average exit quality of the new core were kept equal to the reference values. Thus the power uprate is accommodated by an increase of the core mass flow rate. The findings are as follows: - VIPRE subchannel analysis suggests that, due to its higher fuel to coolant heat transfer area and coolant flow area, the large assembly can operate at a power density 20% higher than the traditional assembly while maintaining the same margin to dryout. - CASMO 2D neutronic analysis indicates that the large assembly can sustain an 18-month irradiation cycle (at uprated power) with 3-batch refueling, <5wt% enrichment with <60 MWD/kg average discharge burnup. Also, the void and fuel temperature reactivity coefficients are both negative and close to those of the traditional BWR core. - The susceptibility of the large assembly core to thermalhydraulic/neutronic oscillations of the density-wave type was explored with an in-house code.
(cont.) It was found that, while well within regulatory limits, the flow oscillation decay ratio of the large assembly core is higher than that of the traditional assembly core. The higher core wide decay ratio of the large assembly core is due to its somewhat higher (more negative) void reactivity coefficient. The pressure drop in the uprated core is 17 %Vo higher than in the reference core, and the flow is 20% higher; therefore, larger pumps will be needed. FRAPCON analysis suggests that the thermo-mechanical performance (e.g., fuel temperature, fission gas release, hoop stress and strain, clad oxidation) of the fuel pins in the large assembly is similar to that of the reference assembly fuel pins. A conceptual mechanical design of the large fuel assembly and its supporting structure was developed. It was found that the water rods and lower tie plate can be used as the main structural element of the assembly, with horizontal support being provided by the top fuel guide plate and core plate assembly, and vertical support being provided by the fuel support duct, which also supports the finger-type control rods.
by Aydin Karahan.
S.M.
Gajev, Ivan. "Sensitivity and Uncertainty Analysis of BWR Stability." Licentiate thesis, KTH, Kärnkraftsäkerhet, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-26387.
Повний текст джерелаQC 20101126
Melara, San Román José. "PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR." Doctoral thesis, Universitat Politècnica de València, 2016. http://hdl.handle.net/10251/61307.
Повний текст джерела[ES] Las oscilaciones de potencia y caudal en un BWR no son deseables. Una de las principales preocupaciones es asegurar, durante oscilaciones de potencia, el cumplimiento de la GDC 10 y 12. GDC 10 requiere que el núcleo del reactor se haya diseñado con un margen adecuado para asegurar que los límites admisibles establecidos en el diseño del combustible no se excederán en cualquier condición de operación normal, incluyendo los efectos de los sucesos operacionales anticipados. GDC 12 requiere garantías de que las oscilaciones de potencia que pueden resultar en condiciones que excedan los límites admisibles establecidos de diseño del combustible, o bien no son posibles o puedan ser detectadas y suprimidas de forma pronta y segura. Si la amplitud de la oscilación es grande, antes de que se produzca el scram las varillas de combustible pueden experimentar secados y remojados periódicos, o si las oscilaciones son suficientemente grandes, un secado extendido. La tasa de amortiguamiento (DR) es la típica figura de mérito de la estabilidad lineal. Para la estimación analítica de la DR los códigos en el dominio de la frecuencia son muy usados. Este tipo de códigos son muy rápidos y sus resultados son muy robustos en comparación con los códigos en el domino temporal, cuyos resultados pueden depender del esquema numérico y la nodalización. El único inconveniente de los códigos en el dominio de la frecuencia es que está limitado al dominio lineal; sin embargo, como los requerimientos regulatorios impuestos por el GDC-12, los reactores deben permanecer estables y, por lo tanto, los reactores deben operar siempre en el dominio lineal. LAPUR es un código de estabilidad en el dominio de la frecuencia que contiene una descripción matemática del núcleo de un reactor de agua en ebullición. Resuelve las ecuaciones de conservación en estado estacionario para el refrigerante y el combustible, las ecuaciones dinámicas para el refrigerante, el combustible y el campo neutrónico en el dominio de la frecuencia. Se han realizado varias mejoras a la versión actual del código, LAPUR 5, con el fin de actualizarlo para su uso con los nuevos tipos de diseño de combustible. La geometría del canal se ha cambiado, el área ha pasado de ser constante a poder considerar área variable. El cálculo de las pérdidas locales debido a los espaciadores y contracciones a lo largo del camino que sigue el flujo se han actualizado, pasando a utilizar correlaciones estándar de la industria. Esta nueva versión del código se ha denominado LAPUR 6. En este trabajo, con el fin de verificar la correcta implementación de estos cambios, se ha realizado una doble validación del código LAPUR 6: En primer lugar se ha realizado una validación exhaustiva de los modelos implementados, comparando los valores de salida de LAPUR 6 para un canal con los resultados de SIMULATE-3. Los modelos termohidráulicos de la CN Cofrentes de SIMULATE-3 han sido validados de forma independiente con los datos experimentales. En segundo lugar se ha desarrollado una metodología para el cálculo de la tasa de amortiguamiento con LAPUR 6, definiendo una matriz de validación de los valores de tasa de amortiguamiento analíticos con valores medidos en la planta. Las tasas de amortiguamiento medidos en la Central Nuclear de Cofrentes tienen valores inferiores al 0.3, confirmando el gran margen de estabilidad de la Central Nuclear de Cofrentes cuando se siguen los procedimiento de operación adecuados, y la comparación con los resultados de LAPUR muestra desviaciones de menos de +/- 0.1. La experiencia acumulada sugiere que la incertidumbre para los rangos bajos de tasas de amortiguamiento es generalmente más grande que para los valores altos. Por último se ha utilizado un generador de señales BWR para la estimación de la incertidumbre de los métodos de análisis de señales utilizados en este trabajo para la estimación experimental de la DR, a partir de la funci
[CAT] Les oscil·lacions de potència i flux en un BWR són molt poc desitjades. Una de les majors preocupacions és assegurar-se, durant les oscil·lacions de potència, del compliment de GDC 10 i 12. GDC 10 requerix que el nucli del reactor estiga dissenyat amb un marge apropiat per a assegurar que els limits admissibles establerts en el disseny del combustible no siguen superats davall cap condició d'operació normal, incloent els incidents esperats d'operació. GDC 12 requerix assegurar que les oscil·lacions de potència que poden resultar en condicions on es superen els limits admissibles establerts en el disseny del combustible no siguen possibles o puguen ser detectades de manera segura e immediata i suprimides. Si l'amplitud de les oscil·lacions és gran, abans que el scram ocórrega les barres experimenten un assecat i remullat periòdic, o si l'oscil·lació és prou gran, un assecat estés. La taxa d'amortiment (DR) és la típica figura de mèrit de l'estabilitat lineal. Per a l'estimació analítica de la DR són molt usats els codis en el domini de la freqüència. Este tipus de codis són molt ràpids i els seus resultats són molt robustos en comparació amb els codis en el domini temporal, els resultats del qual són molt dependents de l'esquema numèric i la nodalizació. L'únic inconvenient del domini de la freqüència és que està limitat al domini lineal, no obstant això, com els requeriments reguladors imposats pel GDC-12, els reactors han de mantener-se estables i, per tant, els reactors han d'operar sempre en el domini lineal. LAPUR és un codi d'estabilitat en el domini de la freqüència que conté una descripció matemàtica del nucli d'un reactor d'aigua en ebullició. Resol les equacions de govern estacionàries del refrigerant i el combustible, les equacions dinàmiques del refrigerant, el combustible i el camp neutrònic en el domini de la freqüència. S'han realitzat diverses millores a la versió anterior del codi, LAPUR 5, amb l'objectiu d'actualitzar-ho per al seu ús amb nous tipus de disseny de combustibles. La geometria del canal s'ha canviat d'àrea constant a variable. Les pèrdues locals degudes als espaciadors i contraccions al llarg del camí del flux s'han actualitzat per a utilitzar correlacions estàndard de la indústria. Esta nova versió és LAPUR 6. En este treball, amb l'objectiu de comprovar la correcta implementació d'estos canvis, s'ha realitzat una doble validació del LAPUR 6: Primer, s'ha realitzat una validació exhaustiva dels models implementats, comparant els valors d'eixida per a un canal de LAPUR 6 amb els resultats de SIMULATE-3. Els models termohidraúlics per a SIMULATE-3 de la Central Nuclear de Cofrentes s'han validat independentment amb dades experimentals. Segon, s'ha desenrotllat una Metodologia per al càlcul de la Taxa d'Amortiment amb LAPUR 6, definint una matriu de validació amb valors de taxes d'amortiment analítics i mesurats en la planta. Anàlisis de les dades mesurades en la Central Nuclear de Cofrentes mostren valors de les taxes d'amortiment inferiors al 0.3, confirmant el gran marge d'estabilitat de la Central Nuclear de Cofrentes quan se seguix un adequat procediment d'operació, i la comparació amb LAPUR mostra desviacions inferiors al +/- 0.1. L'experiència acumulada mostra que la incertesa en el rang de taxes d'amortiment baixes és normalment major que per a valors alts de les taxes d'amortiment. Finalment s'ha utilitzat un generador de senyals per a estimar la incertesa dels mètodes d'anàlisi del senyal utilitzats en este treball per a l'estimació experimental de la taxa d'amortiment emprant la funció d'autocorrelació dels senyals de potència APRM o LPRM.
Melara San Román, J. (2016). PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/61307
TESIS
Hu, Rui Ph D. Massachusetts Institute of Technology. "Stability analysis of natural circulation in BWRs at high pressure conditions." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/46431.
Повний текст джерелаIncludes bibliographical references (leaves 112-115).
At rated conditions, a natural circulation boiling water reactor (NCBWR) depends completely on buoyancy to remove heat from the reactor core. This raises the issue of potential unstable flow. oscillations. The objective of this work is to assess the characteristics of stability in a NCBWR at rated conditions, and the sensitivity to design and operating conditions in comparison to previous BWRs. Two kinds of instabilities, namely Ledinegg flow excursion and Density Wave Oscillations (DWO), have been studied. The DWO analyses were conducted for three oscillation modes: Single Channel thermal-hydraulic stability, coupled neutronics region-wide out-of-phase stability and core-wide in-phase stability. Using frequency domain methods, the three types of DWO stability characteristics of the NCBWR and their sensitivity to the operating parameters and design features have been determined. The characteristic equations are constructed from linearized equations, which are derived for small deviations around steady operating conditions. The Economic Simplified Boiling Water Reactor (ESBWR) is used in our analysis as a reference NCBWR design. It is found that the ESBWR can be stable with a large margin around the operating conditions by proper choice of the core inlet orifice scheme, and for appropriate power to flow ratios. In single channel stability analysis, neutronic feedback is neglected. Design features of the ESBWR, including shorter fuel bundle and use of part-length rods in the assemblies, tend to improve the thermal-hydraulic stability performance. However, the thermal-hydraulic stability margin is still lower than that of a typical BWR at rated conditions. In neutronic-coupled out-of-phase as well as in-phase stability analysis, the perturbation decay ratios for ESBWR at our assumed conditions are higher than that of a typical BWR (Peach Bottom 2) at rated conditions, due to its lower thermal-hydraulic stability margin and higher neutronic feedback.
(cont.) Nevertheless, the stability criteria are satisfied. To evaluate the NCBWR stability performance, comparison with BWR/Peach Bottom 2 at both the rated condition and maximum natural circulation condition has been conducted. Sensitivity studies are performed on the effects of design features and operating parameters, including chimney length, inlet orifice coefficient, power, flow rate, and axial power distribution, reactivity coefficients, fuel pellet-clad gap conductance. It can be concluded that the NCBWR and BWR stabilities are similarly sensitive to operating parameters.
by Rui Hu.
S.M.
Luszczek, Karol. "Validation and Benchmarking of Westinghouse BWR lattice physics methods." Thesis, KTH, Reaktorteknologi, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563.
Повний текст джерелаShirvan, Koroush. "Development of optimized core design and analysis methods for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/80665.
Повний текст джерелаCataloged from PDF version of thesis.
Includes bibliographical references (p. 263-268).
Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between the fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x1 6 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE 14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The optimum core pressure is the same as the current 7.1 MPa. The core exit quality is increased to 19% from the ABWR nominal exit quality of 15%. The pin linear heat generation rate is 20% lower, and the core pressure drop and mass of uranium are 30% lower. The BWR-HD's fuel, modelled with FRAPCON 3.4, showed similar performance to the ABWR pin design. The fuel cycle is only 12 month long, but on the per kWhr, the new design operates with 14% lower fuel cycle front-end costs and similar total fuel cycle cost to the 18 month ABWR fuel cycle. The plant systems outside the vessel are assumed to be the same as the ABWR-1I design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULATE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater temperature was kept the same for the BWR-HD and ABWR which resulted in 4 °K cooler core inlet temperature for the BWR-HD given that its feedwater makes up a larger fraction of total core flow. The stability analysis using the STAB and S3K codes showed satisfactory results for the hot channel, coupled regional out-of-phase and coupled core-wide in-phase modes. A RELAP5 model of the ABWR system was constructed and applied to six transients for the BWR-HD and ABWR. The AMCPRs during all the transients were found to be equal or less for the new design and the core remained covered for both. The lower void coefficient along with smaller core volume proved to be advantages for the simulated transients. Helical Cruciform Fuel (HCF) rods were proposed in prior MIT studies to enhance the fuel surface to volume ratio. In this work, higher fidelity models (e.g. CFD instead of subchannel methods for the hydraulic behaviour) are used to investigate the resolution needed for accurate assessment of the HCF design. For neutronics, conserving the fuel area of cylindrical rods results in a different reactivity level with a lower void coefficient for the HCF design. In single-phase flow, for which experimental results existed, the friction factor is found to be sensitive to HCF geometry and cannot be calculated using current empirical models. A new approach for analysis of flow crisis conditions for HCF rods in the context of Departure from Nucleate Boiling (DNB) and dryout using the two phase interface tracking method was proposed and initial results are presented. It is shown that the twist of the HCF rods promotes detachment of a vapour bubble along the elbows which indicates no possibility for an early DNB for the HCF rods and in fact a potential for a higher DNB heat flux. Under annular flow conditions, it was found that the twist suppressed the liquid film thickness on the HCF rods, at the locations of the highest heat flux, which increases the possibility of reaching early dryout. It was also shown that modeling the 3D heat and stress distribution in the HCF rods is necessary for accurate steady state and transient analyses. The safety analysis of the 20% uprated HCF design in the context of a BWR/4 RPV showed satisfactory AMCHFR performance only if CR is estimated by the EPRI- 1 correlation.
by Koroush Shirvan.
Ph.D.
Conboy, Thomas M. "Thermal-hydraulic analysis of cross-shaped spiral fuel in high power density BWRs." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/41309.
Повний текст джерелаIncludes bibliographical references (p. 199-201).
Preliminary analysis of the cross-shaped spiral (CSS) fuel assembly suggests great thermal-hydraulic upside. According to computational models, the increase in rod surface area, combined with an increase in coolant turbulence and inter-channel mixing will allow for a greater than 25% uprate in total core power, without loss of safety margin. Proper design of the rod dimensions can limit circumferential heat-flux to a peak-to-average ratio of 1.88. Non-uniformities in heat flux due to its unusual geometry seem to particularly ally CSS fuel to the BWR core, where limiting conditions are less likely to be locally influenced. Furthermore, the increase in cooling surface and reduction in central pin thickness is expected to drop fuel centerline temperature an estimated 2000C under nominal operating conditions, a reduction which rises to 3000C at 125% of nominal power conditions. In addition to these advantages, the absence of grid spacers within the CSS fuel assembly is expected to lower pressure losses, aiding natural convection and core stability. Spacers typically account for 25-30% of the total core pressure drop. Experimental measurements of hydraulic: losses for 1.5-meter-long model CSS rods in 4x4 arrays show a larger pressure drop at the same flow velocity than for bare cylindrical rods. However, this results in a CSS-bundle turbulent friction factor which is only 90% of the expected value given its hydraulic diameter. The effect of twist pitch on this pressure drop and friction factor is negligible in the range of twists examined.
(cont.) Combined with the elimination of grid spacers, this results in a 40% reduction in core hydraulic loss from the reference case (neglecting entrance and exit plates). All told, the use of CSS rods should reduce total core pressure drop at nominal power by 9%, in spite of a reduction in core flow area. At 125% of nominal power, this becomes a 16% increase in pressure drop in comparison to the reference core at nominal power.
by Thomas M. Conboy.
S.M.
Gajev, Ivan. "Sensitivity and Uncertainty Analysis of Boiling Water Reactor Stability Simulations." Doctoral thesis, KTH, Kärnkraftsäkerhet, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-105866.
Повний текст джерелаThis work has been preformed thanks to the support of the Swedish Radiation Safety Authority (SSM) and EU project NURISP. QC 20121129
Fritz, Malin. "Control rod drop during hot zero power : RIA in BWR." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-201890.
Повний текст джерелаGuimpelson, Bronislav. "BWR coolant chemistry studies using a recirculating in-pile loop." Thesis, Massachusetts Institute of Technology, 1995. http://hdl.handle.net/1721.1/36949.
Повний текст джерелаNivala, Fernberg Mikael. "BWR In-Core Instrumentation Sensitivity to Material and Geometrical Distortions." Thesis, KTH, Skolan för teknikvetenskap (SCI), 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-188826.
Повний текст джерелаAhnesjö, Magnus. "Tomographic reconstruction of subchannel void measurements of nuclear fuel geometries." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-246288.
Повний текст джерелаHultgren, Ante. "Uncertainty Propagation Analysis for Low Power Transients at the Oskarshamn 3 BWR." Thesis, KTH, Fysik, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-147358.
Повний текст джерелаEllis, Tyler Shawn. "Advanced design concepts for PWR and BWR high-performance annular fuel assemblies." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41268.
Повний текст джерелаIncludes bibliographical references (p. 105-107).
Sobering electricity supply and demand projections, coupled with the current volatility of energy prices, have underscored the seriousness of the challenges which lay ahead for the utility industry. This research addresses the impending global need for electricity through the development of advanced annular fuel designs with both internal and external cooling which can achieve higher power densities and hence, higher electricity output from the same basic reactor vessel and containment. Therefore the objectives of this project are to determine the optimal geometrical design parameters of an annular fuel assembly for both PWRs and BWRs for the purpose of achieving maximum power density. It is theorized that utility companies can utilize this design through either retrofitting of their existing reactor facilities or incorporation of the fuel design into new plant concepts. For the case of annular fuel for PWRs, a high performance uranium nitride fuel assembly concept capable of achieving a 50% higher power density was successfully developed. It is shown that a 5% enriched UN annular-fuel assembly can operate at 150% power density for about 50 effective-full-power-days more than that of the nominal 17xl7 solid-fuel-pin assembly operating at 100% power density. Furthermore, neutronic simulation times of this assembly was reduced from approximately 2 days per simulation for a Monte Carlo based analysis to approximately 2 minutes for a deterministic based simulation via the development of an appropriate correction factor for the CASMO-4 neutron transport code. It was shown that a 25% increase in U238 number density for the un-poisoned pins and a 35% increase for the 10 weight percent gadolinium nitride poisoned pins produced the optimal plutonium tracking and infinite multiplication factor simulation.
(cont.) Finally, the 13x13 annular fuel assembly was shown to have a smaller reactivity swing over the fuel lifetime. Thus it was concluded that an annular uranium nitride assembly at 150% power density can be designed for PWRs so as not to require enrichments above 5% in order to reach the desirable cycle length of 18 months. For the case of annular fuel for BWRs, thermal hydraulic simulations were carried out for a 9x9 solid fuel reference assembly and three different annular assemblies with 5x5, 6x6 and 7x7 fuel pin geometries. Prior research had utilized the Hench-Gillis CPR correlation for all thermal hydraulic simulations and determined that as much as an 11% uprate for 5x5 annular geometries and an 18% uprate for 6x6 annular geometries might be achievable. However, since Hench-Gillis uses bundle average conditions for its calculations, it was theorized that this treatment was not appropriate for annular fuel. A benchmarking analysis against experimental critical power data for a 9x9 assembly confirmed this is a more appropriate heat balance correlation, the EPRI-1 Reddy Fighetti, which was adopted in our simulation of the critical power using the subchannel analysis code VIPRE. Several different strategies were pursued in order to improve the minimum critical heat flux ratio of the three different annular fuel assemblies including optimization of the fuel pin dimensions, fuel pin gap, and orifice loss coefficients. However it was concluded that annular fuel is not a promising strategy for increasing the power density. This can be due to the fact that the CHFR margin gained from the increase in heat transfer surface area is being lost due to the need for increased flow velocity, which retards the CHF for BWR conditions. This is exacerbated by the inability for the coolant in the inner channels to mix with the surrounding subchannels.
by Tyler Shawn Ellis.
S.M.
Outwater, John Ogden. "Design, construction and commissioning of an in-pile BWR coolant chemistry loop." Thesis, Massachusetts Institute of Technology, 1991. http://hdl.handle.net/1721.1/13856.
Повний текст джерелаFridström, Richard. "Response of the Gamma TIP Detectorsin a Nuclear Boiling Water Reactor." Thesis, Uppsala University, Applied Nuclear Physics, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-126969.
Повний текст джерелаIn order to monitor a nuclear boiling water reactor fixed and movable detectors are used, such as the neutron sensitive LPRM (Local Power Range Monitors) detectors and the gamma sensitive TIP (Traversing Incore Probe) detectors. These provide a mean to verify the predictions obtained from core simulators, which are used for planning and following up the reactor operation. The core simulators calculate e.g. the neutron flux and power distribution in the reactor core. The simulators can also simulate the response in the LPRM and TIP detectors. By comparing with measurements the accuracy of the core simulators can be quantified. The core simulators used in this work are PHOENIX4 and POLCA7. Because of the complexity of the calculations, each fuel assembly is divided axially into typically 25 nodes, which are more or less cubic with a side length of about 15 cm. Each axial segment is simulated using a 2D core simulator, in this work PHOENIX4, which provides data to the 3D code, in this case POLCA7, which in turn perform calculations for the whole core. The core simulators currently use both radial pin weights and axial node weights to calculate the gamma TIP detector signal. A need to bring forward new weight factors has now been identified because of the introduction of new fuel designs. Therefore, the gamma TIP detector response has been simulated using a Monte Carlo code called MCNPX for a modern fuel type, SVEA-96 Optima2, which is manufactured by Westinghouse. The new weights showed some significant differences compared to the old weights, which seem to overestimate the radial weight of the closest fuel pins and the axial weight of the node in front of the detector. The new weights were also implemented and tested in the core simulators, but no significant differences could be seen when comparing the simulated detector response using new and old weights to authentic TIP measurements.
Beltran, Arroyos Guillem. "Investigation of Conditions for Activation of Rupture Disk in BWR Containment Filtering System." Thesis, KTH, Kärnkraftsäkerhet, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-45667.
Повний текст джерелаNorberg, Thomas. "Modeling of the steam system in a BWR : A Model of Ringhals 1." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-166821.
Повний текст джерелаAuliano, Manuel. "Investigation and validation of void and pressure drop correlations in BWR fuel assemblies." Thesis, KTH, Fysik, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-169548.
Повний текст джерелаHu, Lin-Wen. "Radiolysis calculations and hydrogen peroxide measurments for the MIT BWR coolant chemistry loop." Thesis, Massachusetts Institute of Technology, 1993. http://hdl.handle.net/1721.1/32590.
Повний текст джерелаGaillard, Mathilde. "Validation of the Westinghouse BWR nodal core simulator POLCA8 against Serpent2 reference results." Thesis, KTH, Fysik, 2021. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-292659.
Повний текст джерелаNär en ny nodal-kärnsimulator utvecklas, som alla andra simulatorer, måste den genomgå en omfattande verifierings och valideringsinsats där den i det första steget kommer att testas mot lämpliga referensverktyg i olika teoretiska riktmärkesproblem. Testserien består av att jämföra flera geometrier, från den enklaste till den mest komplexa, genom att simulera dem med den utvecklade nodkärnsimulatorn och med någon högre ord- ningslösning som representerar referenslösningen, i detta fall på Serpent2 Monte Carlo-transportkoden. Syftet med detta examensarbete är att genomföra en del av dessa tester. Den bestod av att simulera en tredimensionell (3D) 2x2 mini-kokande vattenreaktor (BWR) -kärna med den senaste versionen av Westinghouse BWR- nodalkärnasimulator POLCA8, och att jämföra resultatet av dessa simuleringar mot Serpent2-referensresultat. Före detta arbete testades POLCA8 framgångsrikt på ett 3D-enkanaligt riktmärkesproblem med samma Serpent2 / POLCA8-metodik. Detta riktmärkesproblem som beaktas i detta arbete är dock utmanande i flera aspekter. I själva verket bör nodkärnsimulatorn noggrant förutsäga egenvärdena och kraftfördelningarna mot referensre- sultat, och detta genom att ta hänsyn till axiellt läckage, resulterande från övergången från tvådimensionella (2D) oändliga gitterfysikberäkningar till 3D-simuleringar eller starkt axiellt flöde gradienter på grund av att styrstavarna sätts in eller dras ut efter en viss utarmning. Denna sista effekt är känd som CBH-effekten (Control Blade History) och kommer att vara huvudfokus för denna studie. Förutom utvecklingen av en ny version av nodal core-simulatorn är också en ny version av Westinghouse deterministiska transportkod PHOENIX5 under utveckling. PHOENIX5: s noggrannhet testades indirekt genom detta riktmärke genom att tillhandahålla tvärsnitt för POLCA8-simuleringar. Dessutom genererades Serpent2-baserade nodtvärsnitt till POLCA8 för att tillhandahålla medel för att jämföra dessa två uppsättningar av nodtvärsnittsdata. De erhållna resultaten leder till slutsatsen att CBH-modellen ger mycket bra resultat, särskilt med avseende på alla effektfördelningar, och särskilt de som har tagits bort när man behöver mest.
BOADU, HERBERT ODAME. "CONTINUOUS-TIME OPTIMAL CONTROL OF A SIMULATED BOILING WATER NUCLEAR (BWR) POWER PLANT." Diss., The University of Arizona, 1985. http://hdl.handle.net/10150/188087.
Повний текст джерелаGoronovski, Andrei. "Influence of In-vessel Pressure and Corium Melt Properties on Global Vessel Wall Failure of Nordic-type BWRs." Thesis, KTH, Kärnkraftsäkerhet, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-139534.
Повний текст джерелаAPRI-8
Askari, Behrooz. "An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design /." Zürich : ETH, 2008. http://e-collection.ethbib.ethz.ch/show?type=diss&nr=17720.
Повний текст джерелаLachenmann, Michael [Verfasser], and Hans-Peter [Akademischer Betreuer] Röser. "Missionsanalyse und Nutzlastauswahl des Kleinsatelliten Lunar Mission BW1 / Michael Lachenmann. Betreuer: Hans-Peter Röser." Stuttgart : Universitätsbibliothek der Universität Stuttgart, 2013. http://d-nb.info/1029982236/34.
Повний текст джерелаGulati, Saaransh. "Simulation of liquid entrainment in BWR annular flow using an interface tracking method approach." Thesis, Massachusetts Institute of Technology, 2012. http://hdl.handle.net/1721.1/76966.
Повний текст джерела"June 2012." Cataloged from PDF version of thesis.
Includes bibliographical references (p. 84-90).
by Saaransh Gulati.
S.M.
Inoue, Yuichiro 1969. "Combining thorium with burnable poison for reactivity control of a very long cycle BWR." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/17750.
Повний текст джерелаPage 126 blank.
Includes bibliographical references (p. 104-106).
The effect of utilizing thorium together with gadolinium, erbium, or boron burnable absorber in BWR fuel assemblies for very long cycle is investigated. Nuclear characteristics such as reactivity and power distributions are evaluated using CASMO-4. Without thorium, the results show that gadolinium enriched in Gd-157 has the lowest reactivity swing throughout the cycle. However, the local peaking factor (LPF) in the assembly at beginning-of-life (BOL) is high. The erbium case shows more reactivity swing but the LPF is lowest of all three cases. B4C case has the highest reactivity at BOL which would have to be suppressed by control rods. The most important advantage of B4C over others is the saving of uranium inventory needed to achieve the target exposure of 15 effective full power years (EFPY). Further analysis for transient conditions must be performed to ensure meeting all transient limits. Use of thorium in place of some burnable poison makes it possible to save some uranium enrichment while achieving equivalent discharge burnup to the case without thorium, but only by about 1 %. The benefit is small because almost the same amount of burnable poison is always required for suppressing excess reactivity throughout the cycle. Since Th-232 functions more like U-238 than burnable poison, this limits the allowed thorium to extend discharge burnup. Since all fuel assembly designs in this study have the same target exposure of 15EFPY, the economic performance of each design can be compared based on the amount and enrichment of both uranium and burnable absorbers for each fuel design.
(cont.) The B4C-Al fuel is most economical in overall cost even with large uncertainties. The overall cost of gadolinium and erbium cases are concluded to be about the same when large uncertainties are considered.
by Yuichiro Inoue.
S.M.
Tuvelid, Anna. "Comparison of MELCOR and MAAP calculations of core relocation phenomena in Nordic BWR´s." Thesis, KTH, Fysik, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-194199.
Повний текст джерелаLobdell, John Llewellyn. "Dose rate and spectral photon measurements around a loarge BWR using a tissue equivalent plastic scintillator." Diss., Georgia Institute of Technology, 1995. http://hdl.handle.net/1853/15861.
Повний текст джерелаFeng, Tao. "Measurements on stress corrosion crack initiation for A533B steel in BWR water using tapered tensile specimens." Thesis, University of Newcastle Upon Tyne, 1997. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.388128.
Повний текст джерелаGupta, Atul. "Development of Boiling Water Reactor Nuclear Power Plant Simulator for Human Reliability Analysis Education and Research." The Ohio State University, 2013. http://rave.ohiolink.edu/etdc/view?acc_num=osu1355347881.
Повний текст джерелаCastellanos, Alvarez Larisa. "Application of sub-channel thermal-hydraulic analysis to core calculations with POLCA8 and VIPRE-W." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-393517.
Повний текст джерелаAndrews, Nathan Christopher Ivanov Kostadin N. "Primary calculation of the linear heat rate generation of a BWR pin in the ATR B-11 position." [University Park, Pa.] : Pennsylvania State University, 2010. http://honors.libraries.psu.edu/theses/approved/WorldWideIndex/EHT-238/index.html.
Повний текст джерелаTorregrosa, Martin Claudio. "Coupled 3D Thermo-mechanical Analysis of Nordic BWR Lower Head Failure in case of Core Melt Severe Accident." Thesis, KTH, Fysik, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-141381.
Повний текст джерелаOhlsson, Daniel. "Kartläggning av ventiler innehållande Stellite i reaktornära vattensystem på Forsmark 2." Thesis, Högskolan i Gävle, Avdelningen för Industriell utveckling, IT och Samhällsbyggnad, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:hig:diva-25130.
Повний текст джерелаVid 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.
Loberg, John. "Novel Diagnostics and Computational Methods of Neutron Fluxes in Boiling Water Reactors." Doctoral thesis, Uppsala universitet, Tillämpad kärnfysik, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-133238.
Повний текст джерелаFelaktigt tryckt som Digital Comprehensive Summaries of Uppsala Dissertations from the Faculty of Science and Technology 715
Zakova, Jitka. "Advanced fuels for thermal spectrum reactors." Doctoral thesis, KTH, Reaktorfysik, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-103085.
Повний текст джерелаQC 20121004
Breijder, Paul. "Analysis of Advanced Fuel Behaviour during Loss of Coolant Accident in Swedish Boiling Water Reactor." Thesis, KTH, Kärnkraftsäkerhet, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-44484.
Повний текст джерелаJohnsson, John. "Detailed B-10 depletion in control rods operatingin a Nuclear Boiling Water Reactor." Thesis, Uppsala universitet, Institutionen för materialkemi, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-155416.
Повний текст джерелаHerbst, Matthias G. J. "Effect of chloride on environmentally assisted cracking of low alloy steels in oxygenated high temperature water." Thesis, Liverpool John Moores University, 2014. http://researchonline.ljmu.ac.uk/4569/.
Повний текст джерелаYounkin, Timothy R. "Piecewise prediction of nuclide densities with control blade use as a function of burnup in BWR used nuclear fuel." Thesis, Georgia Institute of Technology, 2014. http://hdl.handle.net/1853/53118.
Повний текст джерелаAl-Ani, Jonathan. "Development of a Nordic BWR plant model in APROS and design of a power controller using the control rods." Thesis, KTH, Fysik, 2021. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-289560.
Повний текст джерелаInom ramen för examensarbete har en indatafil (modell) av en nordisk kokvattenreaktor, BWR, utvecklats i simuleringsverktyget APROS. Anläggningsmodellen är främst utformad för att simulera diskreta effektnivåer och innehåller viktiga system och termohydrauliska komponenter som ingår i ångcykeln, inklusive instrumenterings- och kontrollutrustning (dvs. effekt-, tryck-, nivå- och flödesreglering). Fokus har lagts särskilt på att få till en bra representation av ångcykeln, vilket är avgörande för analys av vatten- och ångbeteendet och dess påverkan på reaktoreffekten. Modellen kan främst användas för simulering av jämviktstillstånd vid full effektdrift och till en viss grad även reducerad effektdrift.
Hu, Chih-Chieh. "Mechanistic modeling of evaporating thin liquid film instability on a bwr fuel rod with parallel and cross vapor flow." Diss., Atlanta, Ga. : Georgia Institute of Technology, 2009. http://hdl.handle.net/1853/28148.
Повний текст джерелаCommittee Chair: Abdel-Khalik, Said; Committee Member: Ammar, Mostafa H.; Committee Member: Ghiaasiaan, S. Mostafa; Committee Member: Hertel, Nolan E.; Committee Member: Liu, Yingjie.
Zinzani, Filippo. "Calculation of the eigenfunctions of the two-group neutron diffusion equation and application to modal decomposition of BWR instabilities." Bachelor's thesis, Alma Mater Studiorum - Università di Bologna, 2007. http://amslaurea.unibo.it/594/.
Повний текст джерелаSkoog, Erik. "CFD Annular Flow Modelling Based on a Three-Field Approach." Thesis, Luleå tekniska universitet, Institutionen för teknikvetenskap och matematik, 2020. http://urn.kb.se/resolve?urn=urn:nbn:se:ltu:diva-80165.
Повний текст джерела