The aim of this master thesis work was to investigate how the uncertainties in nuclear data for 238U affects the uncertainty of keff in criticality simulations for nuclear fuel storages. This was performed by using the Total Monte Carlo (TMC) method which allows propagation of nuclear data uncertainties from basic nuclear physics to reactor parameters, such as keff. The TMC approach relies on simulations with hundreds of calculations of keff with different random nuclear data libraries for 238U for each calculation. The result is a probability distribution for keff where the standard deviation for the distribution represents a spread in keff due to statistical and nuclear data uncertainties. Simulations were performed with MCNP for a nuclear fuel storage representing two different cases:  Normal Case and Worst Case. Normal Case represents a scenario during normal conditions and Worst Case represents accident conditions where optimal moderation occurs. In order to validate the MCNP calculations and the libraries produced with TMC, criticality benchmarks were used. The calculated mean value of keff for the criticality benchmark simulations with random libraries produced with TMC obtained a good agreement with the experimental keff for the benchmarks. This indicates that the libraries used in this this work were of good quality. The TMC method´s drawback is the long calculation time, therefore the new method, fast TMC, was tested.  Both fast TMC and original TMC were applied to the Normal Case. The two methods obtained similar results, indicating that fast TMC is a good option in order to reduce the computational time. The computer time using fast TMC was found to be significantly faster compared with original TMC in this work. The 238U nuclear data uncertainty was obtained to be 209 pcm for the Normal Case, both for original and fast TMC. For the Worst Case simulation the 238U nuclear data uncertainty was obtained to be 672 pcm with fast TMC. These results show the importance of handling uncertainties in nuclear data in order to improve the knowledge about the uncertainties for criticality calculations of keff. / Nukleära databibliotek innehåller all nödvändig information för att till exempel kunna simulera en reaktor eller en bränslebassäng för kärnbränsle. Dessa bibliotek är centrala vid beräkningar av olika reaktorparametrar som krävs för en säker kärnkraftsproduktion. En viktig reaktorparameter är multiplikationskonstanten (keff) som anger reaktiviteten för ett system. Ett kritiskt system (keff = 1) innebär att en kedjereaktion av kärnklyvningar kan upprätthållas. Detta tillstånd erfordras i en reaktor för att möjliggöra elproduktion. I en bränslebassäng där använt kärnbränsle förvaras är det viktigt att systemet är underkritiskt (keff < 1). Olika reaktorkoder används för att utföra dessa beräkningar av keff, vars resultat används i processen för att designa säkra bränsleförråd för kärnbränsle. Dagens nukleära databibliotek innehåller osäkerheter som i sin tur beror på osäkerheter i de modellparametrar som används vid framställningen av biblioteken.  Ofta är dessa nukleära data osäkerheter okända, vilket ger upphov till okända osäkerheter vid beräkning av keff. Vattenfall Nuclear Fuel AB undersöker idag möjligheten att öka anrikningen på bränslet för att minska antalet behövda bränsleknippen för en viss energimängd.  Varje bränsleknippe blir då mer reaktiv och i och med det minskar marginalen till kriticitet i bränslebassängen. Därmed är osäkerheterna för nukleära data viktiga i processen för att kunna beräkna den maximalt tillåtna anrikningen för bränslet. För att undersöka hur stora dessa osäkerheter är, användes en relativ ny metod TMC (Total Monte Carlo) som propagerar osäkerheter i nukleära data till olika reaktorparametrar (t.ex. keff) i en enda simuleringsprocess.  TMC metoden användes för att undersöka hur osäkerheterna i nukleära data för 238U påverkar beräkningar av keff för en bränslebassäng med använt kärnbränsle. Beräkningar utfördes för en bränslebassäng under normala driftförhållanden samt för en olyckshändelse då optimal moderering förekommer. Resultaten visade på att standardavvikelsen för nukleära data för 238U var 209 pcm vid normala driftförhållanden och 672 pcm för fallet med optimal moderering. Den ursprungliga TMC metoden är en tidskrävande metod och nyligen har en snabbare variant av TMC utvecklats. Denna nya metod applicerades också på bränslebassängen under normala driftförhållanden och resultaten jämfördes. Resultaten visade att båda metoderna beräknade samma nukleära dataosäkerhet för 238U och genom att använda den snabba TMC metoden, minskade beräkningstiden betydligt jämfört med att använda den ursprungliga TMC metoden.

For the successful deployment of advanced nuclear systems and optimization of current reactor designs, high quality nuclear data are required. Before nuclear data can be used in applications they must first be evaluated, tested and validated against a set of integral experiments, and then converted into formats usable for applications. The evaluation process in the past was usually done by using differential experimental data which was then complemented with nuclear model calculations. This trend is fast changing due to the increase in computational power and tremendous improvements in nuclear reaction models over the last decade. Since these models have uncertain inputs, they are normally calibrated using experimental data. However, these experiments are themselves not exact. Therefore, the calculated quantities of model codes such as cross sections and angular distributions contain uncertainties. Since nuclear data are used in reactor transport codes as input for simulations, the output of transport codes contain uncertainties due to these data as well. Quantifying these uncertainties is important for setting safety margins; for providing confidence in the interpretation of results; and for deciding where additional efforts are needed to reduce these uncertainties. Also, regulatory bodies are now moving away from conservative evaluations to best estimate calculations that are accompanied by uncertainty evaluations. In this work, the Total Monte Carlo (TMC) method was applied to study the impact of nuclear data uncertainties from basic physics to macroscopic reactor parameters for the European Lead Cooled Training Reactor (ELECTRA). As part of the work, nuclear data uncertainties of actinides in the fuel, lead isotopes within the coolant, and some structural materials have been investigated. In the case of the lead coolant it was observed that the uncertainty in the keff and the coolant void worth (except in the case of 204Pb), were large, with the most significant contribution coming from 208Pb. New 208Pb and 206Pb random nuclear data libraries with realistic central values have been produced as part of this work. Also, a correlation based sensitivity method was used in this work, to determine parameter - cross section correlations for different isotopes and energy groups. Furthermore, an accept/reject method and a method of assigning file weights based on the likelihood function are proposed for uncertainty reduction using criticality benchmark experiments within the TMC method. It was observed from the study that a significant reduction in nuclear data uncertainty was obtained for some isotopes for ELECTRA after incorporating integral benchmark information. As a further objective of this thesis, a method for selecting benchmark for code validation for specific reactor applications was developed and applied to the ELECTRA reactor. Finally, a method for combining differential experiments and integral benchmark data for nuclear data adjustments is proposed and applied for the adjustment of neutron induced 208Pb nuclear data in the fast energy region.

<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)

Features recently added to the nuclear reaction software TALYS allow the use of the GEF model as a fission fragment generator. GEF generates data for fission fragment yields, total excitation energy (TXE), total kinetic energy (TKE) and individual fragment excitation energies (E*) with their standard deviations through Monte Carlo simulations for TALYS. In this work a framework named McPUFF was developed to couple GEF and TALYS and study the propagation of uncertainties in fission fragment data. The GEF model has a set of 94 parameters which were changed in order to produce perturbed output data. Both GEF and TALYS were modified to allow implementation of the Total Monte Carlo (TMC) method which is a method for handling the propagation of uncertainties throughout the simulation process. The developed framework allows the user to control aspects of the nuclear reaction using a set of input files. It is designed to be fast and memory efficient, performing simulations in parallel and storing all results in an object structure. A demonstration of the framework for the neutron induced fission of 235U, 238U and 239Pu was performed. Randomly perturbed sets of fission fragment data were created by GEF and fed into TALYS for simulation of the evaporation process using a Hauser-Feshbach statistical model. The impact that the perturbation of parameters in GEF has on results from TALYS were investigated for prompt particle multiplicity and energy. The results showed that a perturbation of parameter values in GEF by 3 percent has significant effects on values for fission observables produced by TALYS. The TALYS results for 235U showed an uncertainty for prompt neutron multiplicity of σn = 0.16 neutrons with an uncertainty for the neutron energy of σϵn = 0.03 MeV. The corresponding values for the uncertainty of the prompt γ-ray multiplicity were σγ = 0.10 γ-rays with an uncertainty for the γ-ray energy of σϵγ = 0.02 MeV. An investigation of how changes in the angular momentum of the fission fragments affects the evaporation process in GEF and TALYS was performed through the perturbation of the GEF parameter Jscaling. The results highlighted the need to scrutinize the handling of angular momentum in TALYS.