Rozprawy doktorskie na temat „BWR1”

Transmutation of transuranic elements is of interest to lower storage unit cost and long-term radiotoxicity. To make use of existing infrastructure, the deployment of Boiling Water Reactors (BWRs) with hafnium cladding and Mixed Oxide (MOX) fuel was proposed, resulting in a hardening of the neutron spectrum. This work tests varying spatial fuel configurations for maximal burn-up, using Serpent, and study their behaviour in common accident scenarios, simulated by a coupled TRACE/PARCS software suite. To this end, we provide a software solution, which serves to transfer Serpent output of a user defined system in a cross section parameter file, readable by TRACE/PARCS. The results of the transfer were tested for safety performance and, if they provided satisfactory steady states, subjected to a turbine-trip event without bypass, with or without control rod SCRAM. Building on the works by Suvdantstseg [12] and Wallenius & Westlen [7], we chose a Transuranium (TRU) content of 16.48% and a Hafnium-content of 5% with various Higher Actinides (HA) contents and z-axis distributions, intended to either maximize safety performance or minimize void worth and study the results. The chosen fuel loading allows a safe shut-down for both accident scenarios. Sharply rising pressure inside the reactor vessel causes a void collapse. The TRU-content lowers the positive reactivity contribution of increased moderator density, compared to the Uranium Oxide (UOX) baseline. Nonetheless, using a Hf-content of 5% in the cladding and MOX-fuel with 16.48 TRU and 2.06 HA, the void coefficient stays negative during a transitional period of the shutdown, lasting for approximately 200 seconds, before before changing it’s sign.

Includes bibliographical references (p. 205-208).
Thesis (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.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, February 2005.
Includes 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.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2007.
Includes 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.

Best Estimate codes are used for licensing, but with conservative assumptions. It is claimed that the uncertainties are covered by the conservatism of the calculation. As Nuclear Power Plants are applying for power up-rates and life extension, evaluation of the uncertainties could help improve the performance, while staying below the limit of the safety margins.   Given the problem of unstable behavior of Boiling Water Reactors (BWRs), which is known to occur during operation at certain power and flow conditions, it could cause SCRAM and decrease the economic performance of the plant. Performing an uncertainty analysis for BWR stability would give better understating of the phenomenon and it would help to verify and validate (V&V) the codes used to predict the NPP behavior.   This thesis reports an uncertainty study of the impact of Thermal-Hydraulic, Neutronic, and Numerical parameters on the prediction of the stability of the BWR within the framework of OECD Ringhals-1 stability benchmark. The time domain code TRACE/PARCS was used in the analysis. This thesis is divided in two parts: Sensitivity study on Numerical Discretization Parameters (Nodalization, Time Step, etc.) and Uncertainty part.   A Sensitivity study was done for the Numerical Parameters (Nodalization and Time step). This was done by refining all possible components until obtaining Space-Time Converged Solution, i.e. further refinement doesn’t change the solution. When the space-time converged solution was compared to the initial discretization, a much better solution has been obtained for both the stability measures (Decay Ratio and Frequency) with the space-time converged model.   Further on, important Neutronic and Thermal-Hydraulic Parameters were identified and the uncertainty calculation was performed using the Propagation of Input Errors (PIE) methodology. This methodology, also known as the GRS method, has been used because it has been tested and extensively verified by the industry, and because it allows identifying the most influential parameters using the Spearman Rank Correlation.
QC 20101126

[EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. If the oscillation amplitude is large, before the scram occurs the fuel rods may experience periodic dry-out and rewetting, or if the oscillation is larger enough, extended dry-out. The Decay Ratio (DR) is the typical linear stability figure of merit. For analytical estimation of DR frequency domain codes are very useful. These types of codes are very fast and their results are very robust in comparison with time domain codes, whose results may be dependent on numeric scheme and nodalization. The only drawback of frequency domain is that you are limited to the linear domain; however, because of regulatory requirements imposed by GDC-12, reactors must remain stable and, thus, reactors always operate in the linear domain. LAPUR is a frequency domain stability code that contains a mathematical description of the core of a boiling water reactor. It solves the steady state governing equations for the coolant and fuel, and the dynamic equations for the coolant, fuel and the neutron field in the frequency domain. Several improvements have been performed to the current version of the code, LAPUR5, in order to upgrade it for use with new fuel design types. The channel geometry has been changed from constant area to variable area. The local losses due to the spacers and contractions along the flow path have been upgraded to use industry standard correlations. This new version is LAPUR 6. In this work, in order to check the correct implementation of these changes, a two-fold LAPUR 6 validation has been performed: First, an exhaustive validation of the models implemented has been performed, comparing single channels LAPUR 6 outputs against SIMULATE-3 results. Cofrentes NPP SIMULATE-3 thermal-hydraulic models have been independently validated against experimental data. Second, a Methodology for calculating Decay Ratios with LAPUR 6 has been developed, defining a validation matrix against analytical and plant measured decay ratios. Analysis of measured data from the Cofrentes NPP has shown that decay ratios have values lower than 0.3 confirming the large stability margin of Cofrentes NPP when proper operating procedures are followed, and the comparison with LAPUR shows deviations less than +/- 0.1. Past experience suggests that the uncertainty in low decay ratio ranges is usually larger than with higher decay ratio values. Finally a BWR noise generator has been used for estimating the uncertainty of the signal analyses methods used in this work for experimental estimation of decay ratio from the autocorrelation function of the APRM or LPRM power signals.
[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

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007.
Includes 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.

A lattice physics code is a vital tool, forming a base of reactor coreanalysis. It enables the neutronic properties of the fuel assembly to becalculated and generates a proper set of data to be used by a 3-D full coresimulator. Due to advancement and complexity of modern Boiling WaterReactor assembly designs, a new deterministic lattice physics codeis being developed at Westinghouse Sweden AB, namely PHOENIX5.Each time a new code is written, its methodology of solving the neutrontransport equation, has to be validated to make sure it providesreliable output. In a wake of preparation for PHOENIX5 release andconsecutive validation efforts, a set of reference Monte Carlo calculationswas prepared, using the code Serpent. A depletion calculation with achosen type of branch cases was conducted. Methods implemented inPHOENIX5 are based on the Current Coupling Collision Probabilitymethod used in older versions of the code HELIOS. Therefore, a comparisonbetween reference Monte Carlo simulations and HELIOS 1.8.1is made, in order to discover problems inherent to the said method ofsolving the neutron transport equation. A special care should be givenduring PHOENIX5 validation, to issues highlighted in this work.Discrepancies in results of Serpent and HELIOS are attributed mostlyto disparities in the basic nuclear data used by the codes, as well as arange of approximations and corrections adopted by the deterministiccode.Serpent and HELIOS showed a good agreement in a typical voidrange (up to 90 % void) and ‘less’ challenging branches (coolant void,fuel temperature and spacer grid branches). More significant discrepanciesappeared for extreme cases with a very high void and control rodpresence (k1 differences as high as 1000 pcm) and rather pronouncedconcentrations of the natural boron dissolved in coolant (absolute differencesroughly at a level of 900 pcm). The issues do not seem to stemsolely from discrepancies in the nuclear data libraries used by Serpentand HELIOS.Moreover, a coolant void bias was consistently found in the resultsof branch calculation at changing coolant void. This confirms the analogousphenomenon found in previous studies of the CCCP based deterministiccodes. It most probably stems from the assumptions used bythe method while tackling the neutron transport equation, such as theflat source approximation, the isotropic scattering assumption and thetransport correction. An alternative transport correction approximationis proposed to alleviate this issue.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.
Cataloged 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.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007.
Includes 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.

The best estimate codes are used for licensing of Nuclear Power Plants (NPP), but with conservative assumptions. It is claimed that the uncertainties are covered by the conservatism of the calculation. Nowadays, it is possible to estimate certain parameters using non-conservative data with the complement of uncertainty evaluation, and these calculations can also be used for licensing. As NPPs are applying for power up-rates and life extension, new licensing calculations need to be performed. In this case, evaluation of the uncertainties could help improve the performance, while staying below the limit of the safety margins. Given the problem of unstable behavior of Boiling Water Reactors (BWR), which is known to occur at certain power and flow conditions, it could cause SCRAM and decrease the economic performance of the plant. Performing an uncertainty analysis for BWR stability would give better understating of the phenomenon and it would help to verify and validate (V&V) the codes used to predict the NPP behavior. This thesis, reports a sensitivity/uncertainty study of numerical, neutronics, and thermal-hydraulics parameters on the prediction of the BWR stability within the framework of OECD Ringhals-1 (R1 stable reactor) and OECD Oskarshamn-2 (O2 unstable reactor) stability benchmarks. The time domain code TRACE/PARCS was used in the analyses. This thesis is divided in three parts: space-time convergence; uncertainty; sensitivity. A space-time convergence study was done for the numerical parameters (nodalization and time step). This was done by refining nodalization of all components and time step until obtaining space-time converged solution, i.e. further refinement doesn’t change the solution. When the space-time converged solutions were compared to the initial models, much better solution accuracy has been obtained for the stability measures (decay ratio and frequency), for both stable (R1) and unstable (O2) reactors with the space-time converged models. Further on, important neutronics and thermal-hydraulics parameters were identified and an uncertainty calculation was performed using the Propagation of Input Errors (PIE) methodology. This methodology, also known as the GRS method, has been used because it has been extensively tested and verified by the industry, and because it allows identifying the most influential parameters using the spearman rank correlation method. Using the uncertainty method’s results, an attempt has been done to identify the most influential parameters affecting the stability. A methodology using the spearman rank correlation coefficient has been implemented, which helps to identify the most influential parameters on the stability (decay ratio and frequency). Additional sensitivity calculations have been performed for better understanding of BWR stability and parameters that affect it.

This work has been preformed thanks to the support of the Swedish Radiation Safety Authority (SSM) and EU project NURISP. QC 20121129

During operation of nuclear power reactors reactivity initiated accidents (RIA) can occur, such as a control rod drop. If this occurs, the reactivity increase dramatically and leads to an increase in power, fuel enthalpy and fuel temperature. The fuel and reactor can be damaged. A methodology to simulate these accidents has been developed for Forsmark Nuclear Power Plant in cooperation with Westinghouse, referred to as the POLCA7 methodology. The POLCA7 methodology results in a limit for fuel failure regarding reactivity of the control rod that dropped in pcm/control rod percent. The limit is estimated from simulations in POLCA7, a static and deterministic code and POLCA-T, a dynamic code. The aim of this thesis is to evaluate the methodology and investigate what happens in a reactor if a control rod drops during hot zero power. Hot zero power is a phase during start-up, where the power is low (~2% of installed power) and the reactor have operation pressure and temperature. The POLCA7 methodology was applied on historic cycles in Forsmark. To evaluate the POLCA7 methodology the control rod drop was simulated in S3K, a dynamic software. The results from these cycles indicate that the limit for fuel failure set in the POLCA7 methodology in pcm/control rod percent is very conservative for fuel with low and medium burnup. Even though the limit is exceeded, the dynamic simulation in S3K shows that the fuel is far from failure regarding SSM limits in fuel enthalpy and cladding temperature. In this thesis new limits in POLCA7 has been generated, which is remarkably higher than the original limit from the POLCA7 methodology. To challenge the methodology, an unrealistic fuel design was simulated with fuel with high burnup surrounded by high reactive fuel. With this fuel design, the enthalpy limit from SSM was exceeded for the fuel with high burnup. The limit from the POLCA7 methodology was exceeded which indicate that the POLCA7 methodology meets the goal of detecting severe RIAs. Fuel with high burnup seems to be the most important fuel to investigate at a RIA simulation. Another discovery is that POLCA7 gives the most severe accident at 2% power, but in S3K it is given by 3-4% power. This is a problem with the POLCA7 methodology. Suggestions are made on how to lower the calculation time and improve the methodology. A control rod sequence that gives an even power distribution and a core with the fuel with high burnup in the periphery and only a few fresh fuels is preferred to avoid damage at a RIA. A control rod sequence was designed for the new cycle in Forsmark 1, in order to try to create a cycle without problems due to RIA. The new sequence was a success with no control rods exceeding the limit of 82 pcm/control rod percent, and it shows that conclusions about the impact of the sequence are correct. Conclusion is made that the methodology should be further investigated and there are good chances to develop a good and time efficient analysis in the future. One presented suggestion is to have a dynamic simulation of the incident instead of the axial simulation. The evaluation with SSM’s limits would then be direct.

Several BWR utilities are adopting practices for determining the state of control rods based on the neutron sensitive LPRM and gamma sensitive TIP measurement systems of the nuclear reactor core. This method is in this study evaluated by quantitatively analyzing the detector sensitivity to several material and geometrical distortions in the detector vicinity, using the Los Alamos National Laboratory stochastic Monte Carlo code MCNP5 and the Westinghouse 3D core simulator POLCA7. These results are used to determine whether or not there are potential pitfalls that need to be considered when applying the in-core detector based diagnostic methods to determine the state of control rods. This study also addresses the utility concern about shutdown margin deterioration due to potential loss of neutron absorbing material from control rods. It is found that indication of potential control rod leakage by means of the LPRM and the gamma TIP probe detector systems is feasible, provided that both detector systems show similar deviations that correspond to the potential loss of neutron absorbing material from the control rod. Furthermore, it is possible to correlate the observed detector signaldeviation to the magnitude of the distortion. The study lays out criteria that need to befulfilled for indication of control rod leakage to be reasonable, but it is to be noted that nodefinitive stand is or can be taken regarding the state of individual control rods at anyutility. The shutdown margin is found not to be significantly deteriorated in any of the utility cases that are studied under an assumption of control rod loss of neutron absorbing material as apotential cause for anomalous detector readings. In order to provoke a significant effect, asection of the top fifth of the control rod must have lost neutron absorbing material next to a control rod position with already low shutdown margin.

The Westinghouse FRIGG loop in Västerås, Sweden, has been used to study the distribution of steam in the coolant flow of nuclear fuel elements, which is known as the void distribution. For this purpose, electrically heated mock-ups of a quarter BWR fuel bundles in the SVEA-96 geometry were studied by means of gamma tomography in the late 1990s. Several test campaigns were conducted, with good results, but not all the collected data was evaluated at the time. In this work, tomographic raw data of SVEA-96 geometry is evaluated using two different tomographic reconstruction methods, an algebraic (iterative) method and filtered back-projection. Reference objects of known composition (liquid water) are used to quantify the decrease in attenuation arising from the presence of the void, which is used to create a map of the void in the horizontal cross sections of the fuel at various axial locations. The resulting detailed void distributions are averaged over subchannels and the subchannel steam core for comparison with simulations. The focus of this work is on the void distribution at high axial locations in the fuel, in fuel bundles with part-length fuel-rods. Measurements in the region above the part-length rods are compared with simulations and the reliability of each method is discussed. The algebraic method is found to be more reliable than the filtered back-projection method for this setup. A reasonable agreement between measurements and predictions is shown. The void, in both cases, appears to be slightly lower in the corner downstream the part-length rods.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006.
Includes 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.

In 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.

Due to the Three Mile Island accident in 1979 the Swedish government took the decision in 1986 to impose a pressure relief system for Swedish BWR’s which prevents containment overpressure in case of LOCA. This pressure relief system consists of a rupture disks in two different systems, non-filtered system 361 and filtered system 362. During a steam line break it is not clear if an unjustified activation of rupture disk 361 or 362 could possibly occur. If significant amount of nitrogen will leak out from the containment then, there is a risk of low pressure in the containment (e.g. due to activation of containment spray) with leaking rupture disks, which might cause air inflow to the containment and burning of hydrogen, so conditions of activation of rupture disk must be studied. The main objective of this master thesis is the investigation of conditions of activation of rupture disk in BWR containment filtering system. In order to find out these conditions specific software called GOTHIC has been used. The methodology of this master thesis has been modeling different containments with GOTHIC software; this thesis work will go from a simple GOTHIC model, that consist in nine lumped control volumes connected by flow paths, until a more complex GOTHIC model that consist in a combination of lumped and 3D control volumes, connected among them by flow paths and 3D connectors. A large LOCA in the upper part of the reactor vessel will be considerate, due to this severe accident; conditions for the activation of the rupture disk will be complying. It has to be mentioned that pressure in the lumped modeling will be lower than pressure in the 3D volumes. Activation time for the lumped modeling will be 8,5 seconds after the steam break for system 362 and activation time for 3D modeling will be 2,8 seconds for system 362 as well. In neither case 361 system will be activated. Considering this is a nuclear safety study and accuracy must be a key point, for further investigations it might be more than advisable using 3D control volumes instead of lumped control volumes. It has to be mentioned also that due to there is no experimental data, uncertainty regarding to the results exist, and if a further safety analysis want to be done, sensitive study of the parameters implemented on GOTHIC software should be performed in the future.

A nuclear power plant is a very complex dynamic system with a lot of built inregulators and security systems that make it almost impossible to know by reasoning,exactly how the system dynamics is going to react due to e.g. plant modifications,transients or operator behaviors. A common way to find out is to build a computermodel and simulate the system. This master thesis is about building a dynamic modelof the steam system in the boiling water reactor Ringhals 1. The model has beendeveloped in the modeling-/simulating software Dymola and the components arewritten in the programming language Modelica. The model contains the most criticalcomponents in the steam system from reactor tank to condenser and also the mostimportant parts of the control systems. The final model has been compared to real power plant data from Ringhals 1 for fullpower operation, reduced power and a turbine trip. During steady state conditionsthe model has good compliance with the available data in most positions of the steamsystem. Due to absence of good data the results of the dynamic verification for thedrop of load and turbine trip is incomplete. Instead the plausibility of the systembehavior has been done. The results are good but the magnitudes of the transientsare impossible to evaluate. Two major weaknesses have been found during the verification of the model. Theyare the turbine behavior during off-design load and various transients, and the controlof the flow through the tube side of the reheater. The lack of mass flow data is alsomaking it hard to fully trust the model. The final conclusion is that the steam system model is not ready to take on realproblems, but it is a good basis for further development and utilization. The abovementioned problems have to be looked further into and depending on the intendedusage of the model; it may be necessary to modify it to more exactly describe certainparts of the steam system.

When a new nodal core simulator is developed, like all other simulators, it must go through an extensive verification and validation effort where, in the first stage, it will be tested against appropriate reference tools in various theoretical benchmark problems. The series of tests consist of comparing several geometries, from the simplest to the most complex, by simulating them with the nodal core simulator developed and with some higher order solver representing the reference solution, in this case on the Serpent2 Monte Carlo transport code. The aim of this master’s thesis is to carry out one part of these tests. It consisted in simulating a three-dimensional (3D) 2x2 mini boiling water reactor (BWR) core with the latest version of the Westinghouse BWR nodal core simulator POLCA8, and in comparing the outcome of these simulations against Serpent2 reference results. Prior to this work, POLCA8 was successfully tested on a 3D single-channel benchmark problem using the same Serpent2/POLCA8 methodology. However, this benchmark problem considered in this work is challenging in several aspects. Indeed, the nodal core simulator should accurately predict the eigenvalues and power distribu- tions against reference results, and this by taking into account axial leakage, resulting from the passage from two-dimensional (2D) infinite lattice physics calculations to 3D simulations, or strong axial flux gradients due to the insertion or withdrawal of the control rods after a certain depletion. This last effect is known as the Control Blade History (CBH) effect and will be the main focus of this study. In addition to the development of a new version of the nodal core simulator, a new version of the Westinghouse deterministic transport code PHOENIX5 is also under development. The accuracy of PHOENIX5 was indirectly tested through this benchmark by providing the cross sections for the POLCA8 simulations. In addition, Serpent2 based nodal cross sections were generated to POLCA8 to provide means of comparing these two sets of nodal cross section data. The results obtained lead to the conclusion that the CBH model gives very good results, especially with regard to all power distributions, and especially those after the removal of the control bars when needed most.keywords: Nodal Core Analysis, Monte Carlo Methods, CBH Effects
När 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.

A suboptimal controller has been developed for a Boiling Water Reactor Nuclear Power Plant, using the DARE P Continuous Simulation Language, which was developed in the Electrical Engineering Department at the University of Arizona. A set of 48 nonlinear first-order differential equations and a large number of algebraic equations has been linearized about the equilibrium state. Using partitioning, the linearized equations were transformed into a block triangular form. The concept of optimal control and a square performance index reflecting the desired plant behavior have been applied on the slow subsystem to develop a suboptimal controller. The obtained feedback law is shown by simulation to be able to compensate for a variety of plant disturbances. A large variety of responses can be obtained by changing the weighting matrices. The control is basically a regulator approach to speed up response during load demand changes. Several simulations are included to demonstrate the control performance. The variables to be controlled have mainly been the average neutron density and the average coolant temperature. Simplifications have been suggested, thus obtaining considerable savings in the computations and ease in design.

The goal of the present study is to investigate the effect of different scenarios of core degradation in a Nordic-type BWR (boiling water reactor) on the reactor pressure vessel failure mode and timing. Specifically we consider the effects of (i) in-vessel pressure, (ii) melt properties. Control rod guide tube (CRGT) cooling and cooling of the debris from the top are considered as severe accident management (SAM) measures in this study. We also consider the question about minimal amount of debris that can be retained inside the reactor pressure vessel (RPV). Analysis is carried out with coupled (i) Phase-change Effective Convectivity (PECM) model implemented in Fluent for prediction of the debris and melt pool heat transfer, and (ii) structural model of the RPV lower head implemented in ANSYS for simulation of thermo-mechanical creep. The coupling is done through transient thermal load predicted by PECM and applied as a boundary condition in ANSYS analysis. Results of the analysis suggest that applying only CRGT and top cooling is insufficient for maintaining vessel integrity with 0.4 m deep (~12 tons) corium melt pool. The failure of the vessel by thermally induced creep can be expected starting from 5.3 h after the dryout of the debris bed in the lower plenum. However, earlier failure of the instrumentation guide tubes (IGTs) is possible due to melting of the nozzle welding. The internal pressure in the vessel in the range between 3 to 60 bars has no significant influence on the mode and location of the global RPV wall failure. However, depressurization of the vessel can delay RPV wall failure by 46 min for 0.7 m (~ 30 tons) and by 24 min for 1.9 m (~ 200 tons) debris bed. For 0.7 m pool case, changes in vessel pressure from 3 to 60 bars caused changes in liquid melt mass and superheat from ~18 tons at 180 K to ~13 tons at 100 K superheat, respectively. The same changes in pressure for 1.9 m case caused changes in liquid melt mass and superheat from ~40 tons at 42 K to ~10 tons at about 8 K superheat, respectively. Investigation of the influence of melt pool properties on the mode and timing of the vessel failure suggest that the thermo-mechanical creep behavior is most sensitive to the thermal conductivity of solid debris. Both vessel wall and IGT failure timing is strongly dependent on this parameter. For given thermal conductivity of solid debris, an increase in Tsolidus or Tliquidus generally leads to a decrease in liquid melt mass and superheat at the moment of vessel wall failure. Applying models for effective thermal conductivity of porous debris helps to further reduce uncertainty in assessment of the vessel failure and melt ejection mode and timing. Only in an extreme case with Tsolidus, Tliquidus range larger than 600 K, with thermal conductivity of solid 0.5 W∙m‑1∙K‑1 and thermal conductivity of liquid melt 20 W∙m‑1∙K‑1, a noticeable vessel wall ablation and melting of the crust on the wall surface was observed. However, the failure was still caused by creep strain and the location of the failure remained similar to other considered cases.
APRI-8

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012.
"June 2012." Cataloged from PDF version of thesis.
Includes bibliographical references (p. 84-90).
by Saaransh Gulati.
S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.
Page 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.

This report investigates the steps of a one-way coupling between two simulation codes developed by Wesinghouse Electric Sweden AB. The Westinghouse POLCA8 is a three dimensional steady-state diffusion theory code used for simulating the neutronic, thermal and hydraulic behavior of a reactor core. In the  thermal-hydraulic module of the code, each fuel assembly is simulated as a one-dimensional channel, accounting for axial variations of the fuel geometry. While sufficient for many applications, the one-dimensional thermal-hydraulic approach may lack spatial resolution in the case of tilted radial power, very inhomogeneous fuel lattices or for specific calculations such as CHF (Critical Heat Flux) in PWR [3]. This limitation will b avoided by performing a code coupling with the sub-channel analysis code, VIPRE-W, to obtain the radial distribution of thermal-hydraulic parameters for each fuel assembly. In this thesis the codes are one-way coupled . To be able to do a coupling an interface is needed, and this has been created in Matlab. In the interface, the output from POLCA8 is converted into a form suitable to use as an input to VIPRE-W.  As an important first step in the coupling process, I have first analyzed how consistent the codes are when simulating the simplest thermal conditions inside the core. To be able to do the comparison,all values extracted from the sub-channel analysis code VIPRE-W must be converted into assembly-average-values, this is also done in the interface. The thermal-hydraulic parameters that have been  analyzed and compared in the two codes are; mass flux, quality and void.

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.

The focus in this thesis is to improve knowledge of the BWR related uncertainties void, channel bow, and control rods. The presence of void determines the moderation of neutrons in BWRs. A high void fraction is less efficient in moderating neutrons than a low one. As a consequence, the ratio of thermal to fast neutrons is dependent on the surrounding void fraction. In this thesis, calculations with 2D/3D codes corroborate this dependence, the void correlation, to be linear and very robust to changes in different reactor parameters. The void fraction could be predicted from the ratio of simultaneously measured reaction rates from thermal and fast neutron detectors over the whole core with an uncertainty of ±1.5%. The only parameter found disturbing the void correlation significantly is channel bow. However, since channel bow is the only phenomenon found biasing the void correlation, it is found that the void prediction methodology can be used to indicate channel bow with a sensitivity of 4% per mm bow. Consequently, large channel bows could easily be detected. Increased knowledge of void fractions and channel bow could increase both safety and economy of nuclear power production. This thesis also investigates how 2D/3D codes used in production perform in calculating detailed impact of control rods on pin powers and their ability to perform control rod depletion calculations in the reflector region. It is found that the axial resolution used in 3D nodal codes has very large impact on pin power gradients, i.e., using a standard nodal size of ~15 cm can cause underestimations of 50% in pin power gradients, which could lead to fuel damages. In addition, two methods for determining the neutron flux in the control rod when it is withdrawn from the core are presented. Both methods can be used in a 3D nodal code to reproduce the neutron flux in the reflector region with an uncertainty of ±3%.
Felaktigt tryckt som Digital Comprehensive Summaries of Uppsala Dissertations from the Faculty of Science and Technology 715

The advanced fuels investigated in this thesis comprise fuels non− conventional in their design/form (TRISO), their composition (high content of plutonium and minor actinides) or their use in a reactor type, in which they have not been used before (e.g. nitride fuel in BWR). These fuels come with a promise of improved characteristics such as safe, high temperature operation, spent fuel transmutation or fuel cycle extension, for which reasons their potentialis worth assessment and investigation. Their possible use also brings about various challenges, out of which some were addressed in this thesis. TRISO particle fuels with their superior retention abilities enable safe, high−temperature operation. Their combination with molten salt in the Advanced High Temperature Reactor (AHTR) concept moreover promises high operating temperature at low pressure, but it requires a careful selection of the cooling salt and the TRISO dimensions to achieve adequate safety characteristic, incl. a negative feedback to voiding. We show that an AHTR cooled with FLiBe may safely operate with both Pu oxide and enriched U oxide fuels. Pu and Minor Actinides (MA) bearing fuels may be used in BWR for transmutation through multirecycling; however, the allowable amounts of Pu and MA are limited due to the degraded feedback to voiding or low reactivity.We showed that the main positive contribution to the void effect in the fuelswith Pu and MA content of around 11 to 15% consist of the decreased thermalcapture probability in Pu-240, Pu-239 and Am-241 and increased fast and resonance fission probability of U-238, Pu239 and Pu-240. The total void worthmoreover increases during multirecycling, limiting the allowable amount ofMA to 2.45% in uranium−based fuels. An alternative, thorium−based fuel allows for 3.45% MA without entering the positive voiding regime at any point of the multirecycling. The increased alpha−heating associated with the use of transmutation fuels, is at level 24−31 W/kgFUEL in the uranium based fuels and 32−37 W/kgFUEL in the thorium−based configurations. The maximum value of the neutron emission, reached in the last cycle, is 1.7·106 n/s/g and 2·106 n/s/g for uranium and for thorium−based fuels, respectively. Replacing the standard UO2 fuel with higher−uranium density UN orUNZrO2 fuels in BWR shows potential for an increase of the in-core fuelresidence time by about 1.4 year. This implies 1.4% higher availability of the plant. With the nitride fuels, the total void worth increases and the efficiency of the control rods and burnable poison deteriorates, but no major neutronics issue has been identified. The use of nitride fuels in the BWR environment is conditioned by their stability in hot steam. Possible methods for stabilizing nitride fuels in water and steam at 300◦ C were suggested in a recent patentapplication.

QC 20121004

In accident analysis regarding nuclear power plants, it is very common to use thermal hydraulic system codes, such as TRACE, developed by U.S. NRC. In the case of licensing a power plant, this is one of the necessities. TRACE is a relatively new thermal hydraulic system code and a lot of knowledge is needed to implement it in a correct way, especially in accident analysis, where it is a requirement that the rules and statements in Appendix-K, dealing with criteria for ECCS-models, are modelled. In this thesis an improved model of a Swedish Boiling Water Reactor within TRACE is realized and tested. Afterwards, once a working and representative model has been obtained, a sensitivity study in conducted in order to investigate the sensitivity of TRACE for a couple of thermal hydraulic parameters. The sensitivity study is focussing on the eect of the peak cladding temperature, as well as the coolability of the nuclear fuel in terms of quenching and quench-front velocities. It is found to be hard to say unilaterally what the eect of changing a certain number of parameters on the reactor behaviour is. As it turns out to be, although strongly related, the peak cladding temperatures and the quench phenomena can behave dierently

In a nuclear power plant, control rods play a central role to control the reactivity ofthe core. In an inspection campaign of three control rods (CR 99) operated in theKKL reactor in Leibstadt, Switzerland, during 6 respectively 7 consecutive cycles,defects were detected in the top part of the control rods due to swelling caused bydepletion of the neutron-absorbing 10B isotope (Boron-10). In order to correlatethese defects to control rod depletion, the 10B depletion has in this study beencalculated in detail for the absorber pins in the top node of the control rods.Today the core simulator PLOCA7 is used for predicting the behavior of the reactorcore, where the retrievable information from the standard control rod follow-up isthe average 10B depletion for clusters of 19 absorber holes i.e. one axial node.However, the local 10B depletion in an absorber pin may be significantly differentfrom the node average depletion that is re-ceived from POLCA7. To learn more, the 10B depletion has been simulated for each absorber hole in the uppermost node usingthe stochastic Monte Carlo 3D simulation code MCNP as well as an MCNP- based2D-depletion code (McScram). It was found that the 10B depletion is significantly higher for the uppermost absorberpins than the node average. Furthermore, the radial depletion in individual absorberpins was found to be much higher than expected. The results are consistent with theexperimental data on control rod defects.

The aim of this thesis was to derive a better understanding with regard to the effects of chloride on the general corrosion behaviour of low-alloy steels (LAS) in oxygenated high-temperature water (HTW) and to investigate the underlying mechanisms for crack initiation and propaga-tion due to chloride assisted environmentally assisted cracking (EAC). Therefore, systematic investigations on the effect of chloride on the EAC behaviour of LAS were performed to un-derstand and elucidate the underlying mechanisms. The overall thesis is divided into three parts focussing on the effect of chloride on: i) general corrosion, ii) crack initiation, and iii) crack growth of low-alloy steels in oxygenated high-temperature water. Studies on the effect of chloride on the general corrosion behaviour were performed by immer-sion tests that were evaluated using electrochemical monitoring techniques and different post-test investigation methods like SEM, ToF-SIMS, and others. From the performed investiga-tions it is concluded that the presence of small amounts of chloride in oxygenated HTW causes an incorporation of chloride into the oxide layer, a thinning of the oxide layer thickness, and pronounced pitting. The crack initiation susceptibility of LAS was investigated using CERT tests. These tests showed an increased number of crack initiation locations and a decrease of the elongation at fracture with increasing chloride concentrations. Crack growth rate tests clearly demonstrated that not the increase in the chloride concentration per se, but the conjoint occurrence of an active or dormant crack and increased chloride con-centration causes an increase in the observed crack growth rates. For practical applications of LAS in oxygenated HTW this means that short term transients seem to be not harmful regarding component integrity, but long term increased chloride con-centrations should by prohibited since they cause increased general corrosion of LAS. Taking crack initiation and crack growth into consideration, the conjoint occurrence of increased chlo-ride concentrations and mechanical straining at stress levels above the yield strength should be avoided.

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.

In this master thesis an input-model of a Nordic BWR power plant has been developed in APROS. The plant model contains key systems and major thermohydraulic components of the steam cycle, including I&C systems (i.e. power, pressure, level and flow controls). The plant model is primarily designed for balance of plant studies at discrete power levels. The input-model of the power plant focuses especially on the steam cycle which is crucial for analysing water and steam behaviour and its influence on the reactor power. At the current stage, the model primarily handles steady-state conditions of full-power operation, which has been the design point. It has also been shown that reduced-power operation can be simulated with a reasonable trendline of pressure and temperature progression over facility components.
Inom 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.

Thesis (M. S.)--Mechanical Engineering, Georgia Institute of Technology, 2009.
Committee Chair: Abdel-Khalik, Said; Committee Member: Ammar, Mostafa H.; Committee Member: Ghiaasiaan, S. Mostafa; Committee Member: Hertel, Nolan E.; Committee Member: Liu, Yingjie.

In this thesis, numerical methods aiming at determining the eigenfunctions, their adjoint and the corresponding eigenvalues of the two-group neutron diffusion equations representing any heterogeneous system are investigated. First, the classical power iteration method is modified so that the calculation of modes higher than the fundamental mode is possible. Thereafter, the Explicitly-Restarted Arnoldi method, belonging to the class of Krylov subspace methods, is touched upon. Although the modified power iteration method is a computationally-expensive algorithm, its main advantage is its robustness, i.e. the method always converges to the desired eigenfunctions without any need from the user to set up any parameter in the algorithm. On the other hand, the Arnoldi method, which requires some parameters to be defined by the user, is a very efficient method for calculating eigenfunctions of large sparse system of equations with a minimum computational effort. These methods are thereafter used for off-line analysis of the stability of Boiling Water Reactors. Since several oscillation modes are usually excited (global and regional oscillations) when unstable conditions are encountered, the characterization of the stability of the reactor using for instance the Decay Ratio as a stability indicator might be difficult if the contribution from each of the modes are not separated from each other. Such a modal decomposition is applied to a stability test performed at the Swedish Ringhals-1 unit in September 2002, after the use of the Arnoldi method for pre-calculating the different eigenmodes of the neutron flux throughout the reactor. The modal decomposition clearly demonstrates the excitation of both the global and regional oscillations. Furthermore, such oscillations are found to be intermittent with a time-varying phase shift between the first and second azimuthal modes.

This master thesis aim to model the annular flow that occurs in the final section between the fuel rods inside Boiling Water Reactors, by approximating the geometry to a cylindrical pipe. Simulations were performed in the software ANSYS Fluent, as a step in the development of replacing the 1D correlations currently used in the nuclear industry with CFD models in 3D. An Eulerian-Lagrangian approach was used for the three fields of steam, liquid film and liquid droplets in the model. Entrainment was modeled based on 1D correlations from Okawa [7] and deposition with the built in Discrete Phase Model in ANSYS Fluent. The work focused on making the process less time consuming, and increasing accuracy of the model by comparing the results with empirical data based on experimental values. A transverse velocity was applied on the droplets at the point of entrainment with better correlating results with the Okawa model.