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Dissertations / Theses on the topic 'BWR1'

Author: Grafiati

Published: 6 September 2023

Last updated: 31 July 2025

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1

Raub, Sebastian. "Transient behaviour in a BWR with Hafnium Cladding : Feasibility study of using BWRs as Higher Actinide Burners at the Example of Ringhals I." Thesis, KTH, Skolan för teknikvetenskap (SCI), 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-38189.

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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

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2

Chun, John Hwan. "Modeling of BWR water chemistry." Thesis, Massachusetts Institute of Technology, 1990. http://hdl.handle.net/1721.1/13660.

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3

Soma, Kovács István. "Simplified Simulator for BWR Instabilities." Thesis, KTH, Fysik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-210626.

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4

Ferroni, Paolo Ph D. Massachusetts Institute of Technology. "Steady state thermal hydraulic analysis of hydride fueled BWRs." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41263.

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Includes bibliographical references (p. 205-208).<br>Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006.<br>(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 percentag

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5

Morra, Paolo. "Design of annular fuel for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/34448.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, February 2005.<br>Includes bibliographical references (p. 94).<br>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

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6

Karahan, Aydin. "An evolutionary fuel assembly design for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41304.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2007.<br>Includes bibliographical references (p. 138-140).<br>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 dia

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7

Gajev, Ivan. "Sensitivity and Uncertainty Analysis of BWR Stability." Licentiate thesis, KTH, Kärnkraftsäkerhet, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-26387.

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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. Performi

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8

Melara, San Román José. "PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR." Doctoral thesis, Universitat Politècnica de València, 2016. http://hdl.handle.net/10251/61307.

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[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 readil

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9

Hu, Rui Ph D. Massachusetts Institute of Technology. "Stability analysis of natural circulation in BWRs at high pressure conditions." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/46431.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007.<br>Includes bibliographical references (leaves 112-115).<br>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 Ledin

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10

Luszczek, Karol. "Validation and Benchmarking of Westinghouse BWR lattice physics methods." Thesis, KTH, Reaktorteknologi, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563.

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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 PHOENI

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11

Shirvan, Koroush. "Development of optimized core design and analysis methods for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/80665.

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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.<br>Cataloged from PDF version of thesis.<br>Includes bibliographical references (p. 263-268).<br>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 cl

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12

Conboy, Thomas M. "Thermal-hydraulic analysis of cross-shaped spiral fuel in high power density BWRs." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/41309.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007.<br>Includes bibliographical references (p. 199-201).<br>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

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13

Gajev, Ivan. "Sensitivity and Uncertainty Analysis of Boiling Water Reactor Stability Simulations." Doctoral thesis, KTH, Kärnkraftsäkerhet, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-105866.

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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 t

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14

Fritz, Malin. "Control rod drop during hot zero power : RIA in BWR." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-201890.

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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

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15

Guimpelson, Bronislav. "BWR coolant chemistry studies using a recirculating in-pile loop." Thesis, Massachusetts Institute of Technology, 1995. http://hdl.handle.net/1721.1/36949.

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16

Nivala, Fernberg Mikael. "BWR In-Core Instrumentation Sensitivity to Material and Geometrical Distortions." Thesis, KTH, Skolan för teknikvetenskap (SCI), 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-188826.

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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 appl

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17

Ahnesjö, Magnus. "Tomographic reconstruction of subchannel void measurements of nuclear fuel geometries." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-246288.

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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

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18

Hultgren, Ante. "Uncertainty Propagation Analysis for Low Power Transients at the Oskarshamn 3 BWR." Thesis, KTH, Fysik, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-147358.

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19

Ellis, Tyler Shawn. "Advanced design concepts for PWR and BWR high-performance annular fuel assemblies." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41268.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006.<br>Includes bibliographical references (p. 105-107).<br>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 fr

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20

Outwater, John Ogden. "Design, construction and commissioning of an in-pile BWR coolant chemistry loop." Thesis, Massachusetts Institute of Technology, 1991. http://hdl.handle.net/1721.1/13856.

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21

Fridström, Richard. "Response of the Gamma TIP Detectorsin a Nuclear Boiling Water Reactor." Thesis, Uppsala University, Applied Nuclear Physics, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-126969.

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<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 th

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22

Beltran, Arroyos Guillem. "Investigation of Conditions for Activation of Rupture Disk in BWR Containment Filtering System." Thesis, KTH, Kärnkraftsäkerhet, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-45667.

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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 activ

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23

Norberg, Thomas. "Modeling of the steam system in a BWR : A Model of Ringhals 1." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-166821.

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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 Mo

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24

Auliano, Manuel. "Investigation and validation of void and pressure drop correlations in BWR fuel assemblies." Thesis, KTH, Fysik, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-169548.

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25

Hu, Lin-Wen. "Radiolysis calculations and hydrogen peroxide measurments for the MIT BWR coolant chemistry loop." Thesis, Massachusetts Institute of Technology, 1993. http://hdl.handle.net/1721.1/32590.

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26

Gaillard, Mathilde. "Validation of the Westinghouse BWR nodal core simulator POLCA8 against Serpent2 reference results." Thesis, KTH, Fysik, 2021. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-292659.

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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 o

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27

BOADU, HERBERT ODAME. "CONTINUOUS-TIME OPTIMAL CONTROL OF A SIMULATED BOILING WATER NUCLEAR (BWR) POWER PLANT." Diss., The University of Arizona, 1985. http://hdl.handle.net/10150/188087.

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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 slo

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28

Goronovski, Andrei. "Influence of In-vessel Pressure and Corium Melt Properties on Global Vessel Wall Failure of Nordic-type BWRs." Thesis, KTH, Kärnkraftsäkerhet, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-139534.

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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 couple

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29

Askari, Behrooz. "An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design /." Zürich : ETH, 2008. http://e-collection.ethbib.ethz.ch/show?type=diss&nr=17720.

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30

Lachenmann, Michael [Verfasser], and Hans-Peter [Akademischer Betreuer] Röser. "Missionsanalyse und Nutzlastauswahl des Kleinsatelliten Lunar Mission BW1 / Michael Lachenmann. Betreuer: Hans-Peter Röser." Stuttgart : Universitätsbibliothek der Universität Stuttgart, 2013. http://d-nb.info/1029982236/34.

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31

Gulati, Saaransh. "Simulation of liquid entrainment in BWR annular flow using an interface tracking method approach." Thesis, Massachusetts Institute of Technology, 2012. http://hdl.handle.net/1721.1/76966.

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

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32

Inoue, Yuichiro 1969. "Combining thorium with burnable poison for reactivity control of a very long cycle BWR." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/17750.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.<br>Page 126 blank.<br>Includes bibliographical references (p. 104-106).<br>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 begi

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33

Tuvelid, Anna. "Comparison of MELCOR and MAAP calculations of core relocation phenomena in Nordic BWR´s." Thesis, KTH, Fysik, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-194199.

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34

Gupta, Atul. "Development of Boiling Water Reactor Nuclear Power Plant Simulator for Human Reliability Analysis Education and Research." The Ohio State University, 2013. http://rave.ohiolink.edu/etdc/view?acc_num=osu1355347881.

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35

Castellanos, Alvarez Larisa. "Application of sub-channel thermal-hydraulic analysis to core calculations with POLCA8 and VIPRE-W." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-393517.

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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 rad

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36

Lobdell, John Llewellyn. "Dose rate and spectral photon measurements around a loarge BWR using a tissue equivalent plastic scintillator." Diss., Georgia Institute of Technology, 1995. http://hdl.handle.net/1853/15861.

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Feng, Tao. "Measurements on stress corrosion crack initiation for A533B steel in BWR water using tapered tensile specimens." Thesis, University of Newcastle Upon Tyne, 1997. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.388128.

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38

Andrews, Nathan Christopher Ivanov Kostadin N. "Primary calculation of the linear heat rate generation of a BWR pin in the ATR B-11 position." [University Park, Pa.] : Pennsylvania State University, 2010. http://honors.libraries.psu.edu/theses/approved/WorldWideIndex/EHT-238/index.html.

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39

Torregrosa, Martin Claudio. "Coupled 3D Thermo-mechanical Analysis of Nordic BWR Lower Head Failure in case of Core Melt Severe Accident." Thesis, KTH, Fysik, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-141381.

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40

Ohlsson, Daniel. "Kartläggning av ventiler innehållande Stellite i reaktornära vattensystem på Forsmark 2." Thesis, Högskolan i Gävle, Avdelningen för Industriell utveckling, IT och Samhällsbyggnad, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:hig:diva-25130.

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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 cobal

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Loberg, John. "Novel Diagnostics and Computational Methods of Neutron Fluxes in Boiling Water Reactors." Doctoral thesis, Uppsala universitet, Tillämpad kärnfysik, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-133238.

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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 simulta

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42

Zakova, Jitka. "Advanced fuels for thermal spectrum reactors." Doctoral thesis, KTH, Reaktorfysik, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-103085.

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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 th

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43

Breijder, Paul. "Analysis of Advanced Fuel Behaviour during Loss of Coolant Accident in Swedish Boiling Water Reactor." Thesis, KTH, Kärnkraftsäkerhet, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-44484.

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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 tes

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44

Johnsson, John. "Detailed B-10 depletion in control rods operatingin a Nuclear Boiling Water Reactor." Thesis, Uppsala universitet, Institutionen för materialkemi, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-155416.

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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 PLO

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Herbst, Matthias G. J. "Effect of chloride on environmentally assisted cracking of low alloy steels in oxygenated high temperature water." Thesis, Liverpool John Moores University, 2014. http://researchonline.ljmu.ac.uk/4569/.

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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:

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46

Younkin, Timothy R. "Piecewise prediction of nuclide densities with control blade use as a function of burnup in BWR used nuclear fuel." Thesis, Georgia Institute of Technology, 2014. http://hdl.handle.net/1853/53118.

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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 Impleme

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47

Al-Ani, Jonathan. "Development of a Nordic BWR plant model in APROS and design of a power controller using the control rods." Thesis, KTH, Fysik, 2021. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-289560.

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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

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48

Hu, Chih-Chieh. "Mechanistic modeling of evaporating thin liquid film instability on a bwr fuel rod with parallel and cross vapor flow." Diss., Atlanta, Ga. : Georgia Institute of Technology, 2009. http://hdl.handle.net/1853/28148.

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

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Zinzani, Filippo. "Calculation of the eigenfunctions of the two-group neutron diffusion equation and application to modal decomposition of BWR instabilities." Bachelor's thesis, Alma Mater Studiorum - Università di Bologna, 2007. http://amslaurea.unibo.it/594/.

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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 robust

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Skoog, Erik. "CFD Annular Flow Modelling Based on a Three-Field Approach." Thesis, Luleå tekniska universitet, Institutionen för teknikvetenskap och matematik, 2020. http://urn.kb.se/resolve?urn=urn:nbn:se:ltu:diva-80165.

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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 M

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