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4. Results for the Quantification of the Uncertainty on the Secondary Sodium Activation

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4. Results for the Quantification of the Uncertainty on the Secondary Sodium Activation EXAMENSARBETE INOM TEKNISK FYSIK, AVANCERAD NIVÅ, 60 HP STOCKHOLM, SVERIGE 2018 Quantification of the uncertainty on the secondary sodium activation due to uncertainties on nuclear data Master thesis report COPPERE BENJAMIN KTH SKOLAN FÖR ARKITEKTUR OCH SAMHÄLLSBYGGNAD Résumé L’activation du sodium secondaire du cœur ASTRID est une problématique majeure du projet car cette activation nécessite la mise en place de protections neutroniques au niveau du cœur et de l’échangeur intermédiaire. Actuellement, le schéma de calcul fourni des résultats « best estimate », c'est-à-dire sans incertitudes. Pour pouvoir justifier l’utilisation de ces résultats dans les différents projets relatifs à ASTRID, il est nécessaire de connaître l’incertitude sur l’activation du sodium secondaire. Le logiciel NUDUNA a permis d’appliquer une méthode Total Monte-Carlo à notre problème pour déterminer l’incertitude sur l’activation du sodium secondaire. Cette méthode consiste à faire varier les paramètres importants de l’étude de manière aléatoire grâce à des tirages sur les matrices de covariance. Ces tirages aléatoires servent ensuite de données d’entrée au code stochastique MCNP. Après avoir effectué un très grand nombre de calculs MCNP, le principe de Wilks permet de déterminer l’incertitude sur l’activation du sodium secondaire due à l’incertitude sur les données nucléaires. L’application de cette méthode sur ASTRID et Superphénix permet d’aboutir à une valeur d’incertitude d’activation du sodium secondaire convergée. Cette incertitude est de 100% pour le réacteur ASTRID alors que l’incertitude sur l’activation du sodium secondaire est plus faible pour Superphénix avec 66%. L’incertitude due au spectre des neutrons est 9%, valeur plus faible comparée à l’impact des sections efficaces. L’origine de l’incertitude sur l’activation du sodium secondaire provient de l’incertitude sur la section efficace de diffusion élastique du 23Na. La comparaison entre le calcul et la mesure sur le réacteur Superphénix a prouvé que la méthode dans son ensemble est conservative, ce qui est confortant en termes de sureté. Abstract The activation of the sodium secondary circuit in the ASTRID core is a major concern because this activation leads to the setting up of protections on the core and the intermediate heat exchangers. Nowadays, the calculation scheme gives the best estimate values, that is to say, without uncertainties. To justify the use of these values on the different part of the ASTRID project, it is mandatory to evaluate the uncertainty on the activation of the secondary sodium. The software NUDUNA enables to apply a Total Monte-Carlo method which allows determining the uncertainty on the secondary sodium activation. The method consists of varying important parameters of the study by doing random samples on the covariance matrices. These random draws are then used as input to the stochastic code MCNP. After performing many MCNP calculations, the Wilks’s principle enables to determine the uncertainty on the activation of the secondary sodium due to uncertainties on nuclear data. The method is applied on the ASTRID and Superphénix reactors to obtain a converged value of the uncertainty on the activation of the secondary sodium. This uncertainty is 100% for the ASTRID reactor whereas the uncertainty is 66% for Superphénix which is a smaller value. The uncertainty due to the neutron spectrum on the ASTRID activation is 9%. This value is smaller compared to the uncertainty due to neutron cross sections. The origin of the uncertainty on the sodium activation comes from the inelastic scattering cross section of the 23Na nuclide. The comparison between calculations and measurements on the Superphénix reactors proves that the method applies conservatism, which is good in term of safety. COPPERE Benjamin MASTER THESIS REPORT Page: 2/49 Date: 02/2018 FRAMATOME – DTI PLN-F FFP/EP-9013INF/G/fr Acknowledgement My master thesis for the KTH Diploma in Nuclear Engineering took place in the Framatome Company and more precisely the "Neutron transport theory, Radioprotection and Criticality" section in Lyon for a period of 6 months. First of all, I would like to thank Jean-Michel Perrois as Head of the Safety & Processes Department. I would also like to acknowledge Amélie Hee-Duval, head of the "Neutron transport theory, Radioprotection and Criticality" section, for her welcome and her help throughout this internship. I would like to thank my work placement mentors, Guillaume Nolin and Pierre-Marie Demy, for their availability, their help and their sympathies that they gave me during these 6 months. I am grateful to Dr. Oliver Buss for his cooperation and commitment during my master thesis, for his availability and his presence to help me with the NUDUNA software. I have a gratitude for Anne-Claire Scholer, Guillaume Vandermoere, Matthieu Culioli, Florent Beck, Pierre Boisseau, François Mollier and Denis Verrier for their wise advice that allowed me to complete my studies. I especially thank Guillaume Testard with whom I shared my office during these 6 months for his good mood and his precious advice. Finally, I would like to warmly thank all the members of the section for their welcome and good humor on a daily basis. COPPERE Benjamin MASTER THESIS REPORT Page: 3/49 Date: 02/2018 FRAMATOME – DTI PLN-F FFP/EP-9013INF/G/fr Table of contents Résumé ......................................................................................................................................................................... 2 Abstract ........................................................................................................................................................................ 2 Acknowledgement ........................................................................................................................................................ 3 1. Introduction ......................................................................................................................................................... 8 1.1. Context ......................................................................................................................................................... 8 1.2. Choice of tools based on current methods .................................................................................................. 9 2. Problematic of the secondary sodium activation on SFR reactors ...................................................................... 9 2.1. Issues due to the activation of the sodium secondary circuit ...................................................................... 9 2.2. Calculation of the secondary sodium activation ........................................................................................ 10 2.3. ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) .................................... 10 2.4. Superphénix ................................................................................................................................................ 12 3. Methodology and tools used for the evaluation of the uncertainties .............................................................. 13 3.1. Procedures to determine uncertainties due to nuclear data ..................................................................... 13 3.2. Uncertainty propagation theory ................................................................................................................. 14 3.3. Methodology to determine uncertainties due to nuclear data ................................................................. 17 3.3.1. Creation of a nuclide database ........................................................................................................... 17 3.3.2. Variance and covariance information from the ENDF6 file ................................................................ 19 3.3.3. Random draws on the input parameters ........................................................................................... 19 3.3.4. The sum rules ...................................................................................................................................... 21 3.3.5. Creation of the random libraries ........................................................................................................ 24 3.3.6. Stochastic calculations with a transport code .................................................................................... 25 3.3.7. Analysis of the stochastic code results ............................................................................................... 25 3.4. Uncertainty due to the neutron source spectrum ..................................................................................... 27 3.5. Different tools for the study of sodium fast reactors ................................................................................. 28 3.5.1. Stochastic calculation tool: MCNP ...................................................................................................... 28 3.5.2. ADVANTG ............................................................................................................................................ 29 3.5.3. NUDUNA: NUclear Data UNcertainty Analysis ................................................................................... 29 4. Results for the quantification of the uncertainty on the secondary sodium activation .................................... 30 4.1. Procedure to determine uncertainties on the secondary sodium
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