<p>Comprehensive sensitivity and uncertainty analysis has been performed for light-water reactor and heavy-water reactor lattices using three techniques; adjoint-based sensitivity analysis, Monte Carlo sampling, and direct numerical perturbation. The adjoint analysis was performed using a widely accepted, commercially available code, whereas the Monte Carlo sampling and direct numerical perturbation were performed using new codes that were developed as part of this work. Uncertainties associated with fundamental nuclear data accompany evaluated nuclear data libraries in the form of covariance matrices. As nuclear data are important parameters in reactor physics calculations, any associated uncertainty causes a loss of confidence in the calculation results. The quantification of output uncertainties is necessary to adequately establish safety margins of nuclear facilities. In this work, the propagation of uncertainties associated with both physics parameters (e.g. microscopic cross-sections) and lattice model parameters (e.g. material temperature) have been investigated, and the uncertainty of all relevant lattice calculation outputs, including the neutron multiplication constant and few-group, homogenized cross-sections have been quantified. Sensitivity and uncertainty effects arising from the resonance self-shielding of microscopic cross-sections were addressed using a novel set of resonance integral corrections that are derived from perturbations in their infinite-dilution counterparts. It was found that the covariance of the U238 radiative capture cross-section was the dominant contributor to the uncertainties of lattice properties. Also, the uncertainty associated with the prediction of isotope concentrations during burnup is significant, even when uncertainties of fission yields and decay rates were neglected. Such burnup related uncertainties result solely due to the uncertainty of fission and radiative capture rates that arises from physics parameter covariance. The quantified uncertainties of lattice calculation outputs that are described in this work are suitable for use as input uncertainties to subsequent reactor physics calculations, including reactor core analysis employing neutron diffusion theory.</p> / Doctor of Philosophy (PhD)

<p>Simulation plays a key role in nuclear reactor safety analysis and being able to assess the accuracy of results obtained by simulation increases their credibility. This thesis examines the propogation of nuclear data uncertainties through lattice level physics calcualtions. These input uncertainties are in the form of covariance matrices, which dictate the variance and covariance of specified nuclear data to one another. These covariances are available within certain nuclear data libraries, however they are generally only available at infinite dilution for a fixed temperature. The overall goal of this research is to examine the importance of various applications of covariance and their associated nuclear data libraries, and most importantanly to examine the effects of dilution and self-shielding on the results. One source of nuclear data and covariances are the TENDL libraries which are based on a reference ENDF data library and are in continuous energy. Each TENDL library was created by randomly perturbing the reference nuclear data at its most fundamental level according to its covariance. These perturbed nuclear data libraries in TENDL format were obtained and NJOY was used to produce cross sections in 69 groups for which the covariance was calculated at multiple temperatures and dilutions. Temperature was found to have little effect but covarances evaluated at various dilutions did differ significantly. Comparisons of the covariances calculated from TENDL with those in SCALE and ENDF/B-VII also revealed significant differences. The multigroup covariance library produced at this stage was then used in subsequent analyses, along with multigroup covariance libraries available elsewhere, in order to see the differences that arise from covariance library sources. Monte Carlo analysis of a PWR pin cell was performed using the newly created covariance library, a specified reference set of nuclear data, and the lattice physics transport solver DRAGON. The Monte Carlo analysis was then repeated by systematically changing the input covariance matrix (for example using an alternative matrix like that included with the TSUNAMI package) or alternate input reference nuclear data. The uncertainty in k-infinite and the homogenized two group cross sections was assessed for each set of covariance data. It was found that the source of covariance data as well as dilution had a significant effect on the predicted uncertainty in the homogenized cell properties, but the dilution did not significanty affect the predicted uncertainty in k-infinite.</p> / Master of Applied Science (MASc)

McMaster University MASTER OF APPLIED SCIENCES (2017) Hamilton, Ontario (Engineering Physics) TITLE: Comparative study of accident-tolerant fuel for a CANDU lattice AUTHOR: Simon Younan, B.Eng. (McMaster University) SUPERVISOR: Dr. David Novog NUMBER OF PAGES: xiii, 120 / Since the Fukushima accident in 2011, there have been an increasing number of studies on the use of accident-tolerant fuel (ATF) in light water reactors to mitigate the consequences of a future severe accident, by better retaining fission products and/or providing operators more time to implement emergency measures. However, few studies exist for CANDU reactors in this regard. The goal of this study is to determine how different types of ATF are expected to behave in a CANDU lattice when compared to the current UO2 fuels. In particular, this study focuses on neutronic parameters calculated using the Serpent 2 code, but also models heat transfer and stylized accident scenarios. The ATF concepts tested include UO2-SiC composites, UN and UN-based composites, U-9Mo, and fully ceramic microencapsulated (FCM) fuel, along with SiC and SS-coated cladding. Four general conclusions can be drawn: 1. Fuel temperature are lower for ATF as compared to traditional fuels. UO2-SiC composite fuel exhibits a moderate temperature reduction compared to UO2, particularly for fresh fuel. Other ATF fuel materials exhibit a substantial decrease in fuel temperature compared to UO2. The lower fuel temperatures are also accompanied by lower melting temperatures for some fuels, hence each design requires specific assessments on safety. 2. As most ATF have a poorer neutron economy compared to standard fuel designs, enrichment is required to use ATF in a CANDU, particularly for UN and FCM fuel compositions. Coolant void reactivity (CVR) is lowest with FCM fuel and highest with U-9Mo fuel. Fuel temperature coefficient (FTC) is most negative for fuel containing UN or U-9Mo. 3. Changing the cladding material from zircaloy to SiC slightly improves neutron economy, while a FeCrAl surface layer impairs neutron economy. The impact of many ATF sheath materials is to greatly reduce or eliminate hydrogen production in some severe accidents. A specific assessment on hydrogen production was not performed in this study. 4. In stylized accident scenarios, all fuels exhibit only a small temperature spike due to the reactivity insertion of the LOCA as the reactor shutdown limits the power excursion. For cases where Emergency Core Cooling functions as designed, fuel and channel failures are precluded for both traditional fuels and ATF. For cases with impairment of ECC, most ATF fuels show lower fuel temperatures than UO2 fuels and adequate heat removal to the pressure-calandria tube fuel channel. The exception would be Mo-based fuels that reach the melting point prior to establishing an adequately high sheath temperature to sustain radiative heat removal to the PT-CT assembly. / Thesis / Master of Applied Science (MASc) / Since the Fukushima accident in 2011, there have been an increasing number of studies on the use of accident-tolerant fuel in nuclear reactors to mitigate the consequences of a future severe accident, reducing the likelihood and severity of a radiation release. Canadian reactors are of the CANDU design, which differs greatly from the reactors most recent studies have focused on. The goal of this study is to determine the feasibility of using accident-tolerant fuel in CANDU reactors, studying different types. In general, the goal of accident-tolerant fuels in CANDU reactors would be to reduce fuel temperatures and improve fission product retention, reducing the likelihood/magnitude of radioactive releases in a severe accident. However, nearly all types of accident-tolerant fuel would also require the uranium to be slightly enriched as opposed to the current fuel which is based on naturally-occurring uranium. This study outlines the results obtained by computer modelling of accident-tolerant fuel in a CANDU reactor, including the enrichment requirements, changes to important reactivity feedbacks, and impacts on accident performance.

In this work, the antineutrino spectrum as specifically generated by nuclear reactors is studied. The topics covered include corrections and higher-order effects in reactor antineutrino experiments, one of which is covered in Ref. [1] and another contributes to Ref. [2]. In addition, a practical application, antineutrino safeguards for nuclear reactors, as summarized in Ref. [3,4] and Ref. [5], is explored to determine its viability and limits. The work will focus heavily on theory, simulation, and statistical analyses to explain the corrections, their origins, and their sizes, as well as the applications of the antineutrino signal from nuclear reactors. Chapter [1] serves as an introduction to neutrinos. Their origin is briefly covered, along with neutrino properties and some experimental highlights. The next chapter, Chapter [2], will specifically cover antineutrinos as generated in nuclear reactors. In this chapter, the production and detection methods of reactor neutrinos are introduced as well as a discussion of the theories behind determining the antineutrino spectrum. The mathematical formulation of neutrino oscillation will also be introduced and explained. The first half of this work focuses on two corrections to the reactor antineutrino spectrum. These corrections are generated from two specific sources and are thus named the spent nuclear fuel contribution and the non-linear correction for their respective sources. Chapter [3] contains a discussion of the spent fuel contribution. This correction arises from spent nuclear fuel near the reactor site and involves a detailed application of spent fuel to current reactor antineutrino experiments. Chapter [4] will focus on the non-linear correction, which is caused by neutron-captures within the nuclear reactor environment. Its quantification and impact on future antineutrino experiments are discussed. The research projects presented in the second half, Chapter [5], focus on neutrino applications, specifically reactor monitoring. Chapter [5] is a comprehensive examination of the use of antineutrinos as a reactor safeguards mechanism. This chapter will include the theory behind safeguards, the statistical derivation of power and plutonium measurements, the details of reactor simulations, and the future outlook for non-proliferation through antineutrino monitoring. / Ph. D.

Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the sample’s burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope results are indicator signatures of extended shutdown or time away from power. Due to dynamic operation in time and function, detailed power history reconstruction for research reactors is very challenging. Frequent variations in power, repeated variable shutdown time length, and experimentation history affect the spectrum an individual assembly is burned with such that full reactor parameter reconstruction is difficult. The results from this technical nuclear forensic analysis may be used with law enforcement, intelligence data, macroscopic and microscopic sample characteristics in a process called attribution to suggest or exclude possible sources of origin for a sample.

The Monte Carlo method is, practically, the only approach capable of giving detail insight into complex neutron transport problems. In reactor physics, the method has been used mainly for determining the keff in criticality calculations. In the last decade, the continuously growing computer performance has allowed to apply the Monte Carlo method also on simple burnup simulations of nuclear systems. Nevertheless, due to its extensive computational demands the Monte Carlo method is still not used as commonly as deterministic methods. One of the reasons for the large computational demands of Monte Carlo criticality calculations is the necessity to carry out a number of inactive cycles to converge the fission source. This thesis presents a new concept of fission matrix based Monte Carlo criticality calculations where inactive cycles are not required. It is shown that the fission matrix is not sensitive to the errors in the fission source, and can be thus calculated by a Monte Carlo calculation without inactive cycles. All required results, including keff, are then derived via the final fission matrix. The confidence interval for the estimated keff can be conservatively derived from the variance in the fission matrix. This was confirmed by numerical test calculations of Whitesides's ``keff of the world problem'' model where other Monte Carlo methods fail to estimate the confidence interval correctly unless a large number of inactive cycles is simulated.   Another problem is that the existing Monte Carlo criticality codes are not well shaped for parallel computations; they cannot fully utilise the processing power of modern multi-processor computers and computer clusters. This thesis presents a new parallel computing scheme for Monte Carlo criticality calculations based on the fission matrix. The fission matrix is combined over a number of independent parallel simulations, and the final results are derived by means of the fission matrix. This scheme allows for a practically ideal parallel scaling since no communication among the parallel simulations is required, and no inactive cycles need to be simulated.   When the Monte Carlo criticality calculations are sufficiently fast, they will be more commonly applied on complex reactor physics problems, like non-linear steady-state calculations and fuel cycle calculations. This thesis develops an efficient method that introduces thermal-hydraulic and other feedbacks into the numerical model of a power reactor, allowing to carry out a non-linear Monte Carlo analysis of the reactor with steady-state core conditions. The thesis also shows that the major existing Monte Carlo burnup codes use unstable algorithms for coupling the neutronic and burnup calculations; therefore, they cannot be used for fuel cycle calculations. Nevertheless, stable coupling algorithms are known and can be implemented into the future Monte Carlo burnup codes. / QC 20100709

Standardmäßig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte für Reaktorkernrechnungen mit 2D-Zellcodes. Ziel dieser Arbeit war es, einen 3D-Zellcode zu entwickeln, mit diesem Code 3D-Effekte zu untersuchen und die Notwendigkeit einer 3D-Datenaufbereitung der Neutronenwirkungsquerschnitte zu bewerten. Zur Berechnung des Neutronentransports wurde die Methode der Erststoßwahrscheinlichkeiten, die mit der Ray-Tracing-Methode berechnet werden, gewählt. Die mathematischen Algorithmen wurden in den 2D/3D-Zellcode TransRay umgesetzt. Für den Geometrieteil des Programms wurde das Geometriemodul eines Monte-Carlo-Codes genutzt. Das Ray-Tracing in 3D wurde auf Grund der hohen Rechenzeiten parallelisiert. Das Programm TransRay wurde an 2D-Testaufgaben verifiziert. Für einen Druckwasser-Referenzreaktor wurden folgende 3D-Probleme untersucht: Ein teilweise eingetauchter Regelstab und Void (Vakuum oder Dampf) um einen Brennstab als Modell einer Dampfblase. Alle Probleme wurden zum Vergleich auch mit den Programmen HELIOS (2D) und MCNP (3D) nachgerechnet. Die Abhängigkeit des Multiplikationsfaktors und der gemittelten Zweigruppenquerschnitte von der Eintauchtiefe des Regelstabes bzw. von der Höhe der Dampfblase wurden untersucht. Die 3D berechneten Zweigruppenquerschnitte wurden mit drei üblichen Näherungen verglichen: Lineare Interpolation, Interpolation mit Flusswichtung und Homogenisierung. Am 3D-Problem des Regelstabes zeigte sich, dass die Interpolation mit Flusswichtung eine gute Näherung ist. Demnach ist hier eine 3D-Datenaufbereitung nicht notwendig. Beim Testfall des einzelnen Brennstabs, der von Void umgeben ist, erwiesen sich die drei Näherungen für die Zweigruppenquerschnitte als unzureichend. Demnach ist eine 3D-Datenaufbereitung notwendig. Die einzelne Brennstabzelle mit Void kann als der Grenzfall eines Reaktors angesehen werden, in dem sich eine Phasengrenzfläche herausgebildet hat.

Standardmäßig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte für Reaktorkernrechnungen mit 2D-Zellcodes. Ziel dieser Arbeit war es, einen 3D-Zellcode zu entwickeln, mit diesem Code 3D-Effekte zu untersuchen und die Notwendigkeit einer 3D-Datenaufbereitung der Neutronenwirkungsquerschnitte zu bewerten. Zur Berechnung des Neutronentransports wurde die Methode der Erststoßwahrscheinlichkeiten, die mit der Ray-Tracing-Methode berechnet werden, gewählt. Die mathematischen Algorithmen wurden in den 2D/3D-Zellcode TransRay umgesetzt. Für den Geometrieteil des Programms wurde das Geometriemodul eines Monte-Carlo-Codes genutzt. Das Ray-Tracing wurde auf Grund der hohen Rechenzeiten parallelisiert. Das Programm TransRay wurde an 2D-Testaufgaben verifiziert. Für einen Druckwasser-Referenzreaktor wurden folgende 3D-Probleme untersucht: Ein teilweise eingetauchter Regelstab und Void (bzw. Moderator mit geringerer Dichte) um einen Brennstab als Modell einer Dampfblase. Alle Probleme wurden zum Vergleich auch mit den Programmen HELIOS (2D) und MCNP (3D) nachgerechnet. Die Abhängigkeit des Multiplikationsfaktors und der gemittelten Zweigruppenquerschnitte von der Eintauchtiefe des Regelstabes bzw. von der Höhe der Dampfblase wurden untersucht. Die 3D berechneten Zweigruppenquerschnitte wurden mit drei üblichen Näherungen verglichen: linearer Interpolation, Interpolation mit Flusswichtung und Homogenisierung. Am 3D-Problem des Regelstabes zeigte sich, dass die Interpolation mit Flusswichtung eine gute Näherung ist. Demnach ist hier eine 3D-Datenaufbereitung nicht notwendig. Beim Testfall des einzelnen Brennstabs, der von Void (bzw. Moderator geringerer Dichte) umgeben ist, erwiesen sich die drei Näherungen für die Zweigruppenquerschnitte als unzureichend. Demnach ist eine 3D-Datenaufbereitung notwendig. Die einzelne Brennstabzelle mit Void kann als der Grenzfall eines Reaktors angesehen werden, in dem sich eine Phasengrenzfläche herausgebildet hat.

Der Reaktordynamikcodes DYN3D wurde in der neu entwickelten Mehrgruppen-Version DYN3D-MG für die Anwendung auf wassergekühlte Reaktoren alternativ zu industriellen DWR und SWR ertüch-tigt. Es wurde die Anwendbarkeit für den graphitmoderierten Druckröhrenreaktor EGP-6 (KKW Bilibi-no), eine Konzeptstudie eines fortgeschrittenen Siedewasserreaktors mit schnellem Neutronenspekt-rum (RMWR) und das Reaktorkonzept RUTA-70 zur Wärmeversorgung nachgewiesen. Beim RUTA-Reaktor geht es vor allem um die Modellierung des Naturumlaufs des Kühlmittels bei niedrigen Sys-temdrücken. Zur Validierung wurden Experimente zu flashing-induzierten Naturumlaufinstabilitäten an der Versuchsanlage CIRCUS der TU Delft mit RELAP5 nachgerechnet. Für die Anwendung von DYN3D auf die alternativen Reaktorkonzepte wurden Modellerweiterungen und Anpassungen vorgenommen, u.a. Modifikationen in den Wärmeleitungs- und -übergangsmodellen. Vergleichsrechnungen mit dem stationären russischen Feingitter-Diffusionscode ACADEM ergänzen die Verifikationsdatenbasis von DYN3D-MG. Zur Validierung wurden zwei reak-tordynamische Experimente am Reaktor EGP-6 nachgerechnet. Für Reaktoren EGP-6, RMWR und RUTA wurden verschiedene Transienten mit Ausfahren von Re-gelstäben mit und ohne Reaktorschnellabschaltung gerechnet. Weiterhin wurden Analysen für den ATWS-Störfall \"Abschalten aller Hauptkühlmittelpumpen bei Vollleistung\" für den RUTA-Reaktor mit den gekoppelten Programmkomplexen DYN3D/ATHLET und DYN3D/RELAP5 durchgeführt. Der Reaktor geht in einen sicheren Zustand mit reduzierter Leistung bei Naturumlauf des Kühlmittels über. Die Ergebnisse von Analysen zum unkontrollierten Ausfahren einer Regelgruppe für den RMWR lassen dagegen eine belastbare Schlussfolgerung bezüglich der Beherrschbarkeit des Aus-fahrens einer Regelgruppe nicht zu. Abschließend wurde der Nutzen der Programmertüchtigung von DYN3D für die Anwendung auf GenIV -Konzepte und LWR mit hohem Konversionsfaktor bewertet.

Made available in DSpace on 2014-10-09T12:35:04Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:00:44Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP