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.

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&amp;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. / <p>This work has been preformed thanks to the support of the Swedish Radiation Safety Authority (SSM) and EU project NURISP. QC 20121129</p>

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.

<p>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.</p>

This work focuses on both high fidelity experimental and numerical thermal hydraulic studies and advanced frequency decomposition methods. The major contribution of this work is a proposed method for spatio-temporal decomposition of frequencies present in the flow. This method provides an instantaneous visualization of coherent frequency ?structures? in the flow. The significance of this technique from an engineering standpoint is the ease of implementation and the importance of such a tool for design engineers. To validate this method, synthetic verification data, experimental data sets, and numerical results are used. The first experimental work involves flow through the side entry orifice (SEO) of a boiling water reactor (BWR) using non-intrusive particle tracking velocimetry (PTV) techniques. The second experiment is of a simulated double ended guillotine break in the prismatic block gas cooled reactor. Numerical simulations of jet flow mixing in the lower plenum of a prismatic block high temperature gas cooled reactor is used as a final data set for verification purposes as well as demonstration of the applicability of the method for an actual computational fluid dynamics validation case.

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.

[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

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.

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