Advanced non-destructive methods for criticality safety and safeguards of used nuclear fuel

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Advanced non-destructive methods for criticality

safety and safeguards of used nuclear fuel

Riccardo Rossa

Thèse présentée en vue de l’obtention du grade de Docteur en Sciences de l’Ingénieur

Promoteur: Prof. Pierre-Etienne Labeau Co-Promoteur: Prof. Nicolas Pauly

Mentor (SCK•CEN): Dr. Alessandro Borella Co-Mentor (SCK•CEN): Ir. Klaas van der Meer

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Acknowledgements

First of all I would like to express my sincere gratitude to my mentor at SCK•CEN Alessandro Borella that guided me during the Ph.D. research. He taught me a lot about NDA techniques and Monte Carlo simulations and this thesis would not be in this form without his patience, efforts, and dedication.

Remaining in the SPS expert group at SCK•CEN I would like to thank the other colleagues for the nice time spent together. In particular a big thank you to Klaas van der Meer for everything he does for the group both during working hours and with social events.

Special thanks go to my promotor and co-promotor at ULB Pierre-Etienne Labeau and Nicolas Pauly for their advices during those years, and to the rest of the Ph.D. jury: Pierre Capel, Alain Dubus, Paolo Peerani, and Peter Schillebeeckx for their review of the manuscript.

I would like to acknowledge the scientists at JRC-IRMM in Geel working at the GELINA facility for their support during the experimental measurements and data analysis: Peter Schillebeeckx, Carlos Paradela, Jan Heyse, Stefan Kopecky, Ruud Wynants, Gery Alaerts.

Thank you to all the friends that shared these years in the Boeretang Kingdom: I would need another book to thank you all individually! You may be scattered now all over the world, but the nice times together will stay forever in my mind.

Last but not least I would like to thank my family for the continuous support during these years. Switching to my native language, un grazie speciale alla mia famiglia che mi ha continuato a supportare in tutti questi anni.

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List of figures

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Figure 4-10: View of the central guide tube in the Monte Carlo model and comparison of different

positioning of the detector... 45

Figure 4-11: Comparison of detectors with different lengths (L1 and L2). The track of a neutron crossing both detectors is also shown with the indication of the incoming angles ( and ). ... 46

Figure 4-12: Zoom of the central region of the fuel assembly and comparison of the detector cover by the SINRD filter. ... 47

Figure 4-13: Energy distribution of the difference between the detector response calculated for a 239 Pu fission chamber covered by Gd and Cd filters. The fuel in these simulations contained 238U and 16 O. ... 49

Figure 5-1: Aerial view of the GELINA Time-of-Flight facility at the JRC-IRMM in Geel. ... 54

Figure 5-2: Schematic view of a transmission experiment. ... 55

Figure 5-3: Schematic representation of the self-indication experiments carried out at GELINA. ... 56

Figure 5-4: Transmission through different Gd and Cd foils. The experimental transmission is compared with the analytical transmission based on the calculations in Chapter 4. ... 58

Figure 5-5: Transmission through different Gd and Cd foils using different nuclear data libraries. The values were calculated with the analytical approach described in Chapter 4. The plot on the right is focused on energy range below 0.1 eV. ... 59

Figure 5-6: Experimental setup for the self-indication experiments. The 0.027 mm Cd sample is placed in the neutron beam and is surrounded by 4 C6D6 scintillator detectors. ... 60

Figure 5-7: Spectra of the self-indication experiments with Cd samples in the beam. The spectrum obtained with the detector only is reported for comparison together with the background contribution. All spectra were normalized to the same beam intensity. ... 61

Figure 5-8: Spectra of the self-indication experiments with a 0.03 mm Gd (left) and 1.0 mm Cd (right) filter in the beam. The spectrum obtained with the detector only is reported for comparison together with the background contribution. ... 62

Figure 5-9: Experimental observables RSI,1 and RSI,2 as a function of the areal density of the Cd sample placed in the beam. The results were normalized to the measurements without Cd sample. ... 63

Figure 5-10: Spectra obtained for the 235U fission chamber with a 0.03 mm Gd (left) and 1.0 mm Cd (right) filter in the beam. Moreover, several Cd samples were used with the Gd filter to simulate the neutron absorption by fuel pins containing 239Pu. ... 64

Figure 5-11: Spectra obtained for the 10B ionization chamber with a 0.03 mm Gd (left) and 1.0 mm Cd (right) filters in the beam. Moreover, several Cd samples were used with the Gd filter to simulate the neutron absorption by fuel pins containing 239Pu. ... 64

Figure 5-12: Experimental observable RSI,2 as a function of the areal density of the Cd sample placed in the beam. The data refer to measurements with a self-indication detector, a 235U fission chamber, and a 10B ionization chamber. ... 65

Figure 6-1: Importance function for a 235U fission chamber placed in different guide tubes. The neutron flux was calculated in the guide tube depicted in grey. The color bar ranges between 0 and 1%. ... 68

Figure 6-2: Importance function for a 238U fission chamber placed in different guide tubes. The neutron flux was calculated in the guide tube depicted in grey. The color bar ranges between 0 and 1%. ... 69

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Figure 6-4: Areas identified for the calculation of the integral contribution to the detector response in the different guide tubes. ... 70 Figure 6-5: Neutron detector response of 235U fission chambers and 238U fission chambers. The results for each plot were normalized to the maximum value obtained in the correponding simulation, and the uncertainty of the values was lower than 0.2% ... 73 Figure 6-6: Normalized detector responses for different guide tubes. The 235U and 238U fission chambers were compared and the statistical uncertainty of the simulations was also included. The results for the other guide tubes were not included due to the symmetry of the fuel assembly. ... 73 Figure 6-7: Gamma-ray detector response of an ionization chamber with nitrogen as filling gas at 1 atm. The results were normalized to the maximum value obtained in the guide tubes, and the uncertainty of the values was lower than 0.8% ... 74 Figure 6-8: Normalized detector responses for different guide tubes. Only the response of ionization chamber with nitrogen as filling gas at 1 atm was reported because the responses of the other detector types were within the statistical uncertainty. The results for the other guide tubes were not included due to the symmetry of the fuel assembly. ... 75 Figure 6-9: Normalized detector responses for fuel with different burnup (BU). The results refer to 235

U fission chambers and the values for the other guide tubes were not included due to the symmetry of the fuel assembly. ... 76 Figure 6-10: Normalized detector responses for fuel with different burnup (BU). The results refer to 238

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List of tables

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List of acronyms

AP Additional Protocol

ASEA Allmänna Svenska Elektriska Aktiebolaget (general Swedish electric company)

BU BUrnup

BWR Boiling Water Reactor

C/S Containment and Surveillance

CLAB Centralt mellanlager för använt kärnbränsle (central interim storage facility for spent nuclear fuel)

CT Cooling Time

DCVD Digital Cherenkov Viewing Device

DDA Differential Die-Away

DDSI Differential Die-away Self-Interrogation

DG-ENER European Commission's Directorate General for ENERgy

ENDF Evaluated Nuclear Data File

EU European Union

EURATOM EURopean ATOMic energy community

eV electronVolt

GELINA GEel LINear Accelerator

GWd GigaWatt-day

HM Heavy Metal

ICVD Improved Cherenkov Viewing Device

IAEA International Atomic Energy Agency

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IRMM Institute for Reference Materials and Measurements

ITU Institute for TransUranium elements

JANIS JAva-based Nuclear Information Software

JEFF Joint Evaluated Fission and Fusion File

JENDL Japanese Evaluated Nuclear Data Library

JRC Joint Research Centre

LANL Los Alamos National Laboratory

LEU Low Enriched Uranium

LLNL Lawrence Livermore National Laboratory

LWR Light Water Reactor

MATLAB MATrix LABoratory

MCNPX Monte Carlo N-Particle eXtended

MOX Mixed OXide

NDA Non-Destructive Assay

NGSI Next Generation Safeguards Initiative

NGSI-SF Next Generation Safeguards Initiative – Spent Fuel

NMA Nuclear Material Accountancy

NPT Non-Proliferation Treaty

NNWS Non-Nuclear Weapon State

NRD Neutron Resonance Densitometry

NWS Nuclear Weapon State

ORIGEN-ARP Oak Ridge Isotope GENerator – Automatic Rapid Processing

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Pa Pascal

PDET Partial DEfect Tester

ppm part per million

PWR Pressurized Water Reactor

SCALE Standardized Computer Analyses for Licensing Evaluations

SCK•CEN StudieCentrum voor Kernenergie – Centre d'Étude de l'énergie Nucléaire (Belgian nuclear research centre)

SFAT Spent Fuel Attribute Tester

SKB Svensk Kärnbränslehantering Aktiebolag (Swedish nuclear fuel and waste management company)

SINRD Self-Indication Neutron Resonance Densitometry

ToF Time-of-Flight

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Summary

Introduction

This Ph.D. project describes the development of non-destructive assay (NDA) methods for the measurement of spent nuclear fuel and was conducted as a collaboration between the Université Libre de Bruxelles (ULB) and the Belgian nuclear research centre SCK•CEN.

Spent nuclear fuel refers to fuel assemblies that are discharged from nuclear reactors after irradiation and are transferred to an interim storage. This material contains radioactive elements that are responsible for neutron and gamma emissions. Since the radioactive decay leads also to decay heat, the spent fuel is normally stored under water to ensure appropriate cooling and provide radiation shielding.

After irradiation spent fuel still contains about 2% of fissile materials (i.e. 235U and 239Pu), therefore nuclear safeguards are applied to ensure that the material is used only for peaceful applications. The plutonium contained in spent fuel is a major concern for the safeguards community because it represents almost 80% of all material placed under safeguards today. Moreover, the total spent fuel inventory increases with time due to the discharge of fuel assemblies from operating reactors. Several NDA techniques are used for the safeguards verifications of spent fuel and additional techniques are under development to provide more accurate measurements. Both passive and active techniques are considered. Passive techniques rely on the spontaneous emission of radiation from the spent fuel itself, whereas external sources are used for active techniques. The radiation measured by each NDA method varies between neutrons, gamma-rays, and Cherenkov light. The techniques investigated in this Ph.D. project are the Self-Indication Neutron Resonance Densitometry (SINRD) and Partial Defect Tester (PDET).

Development of the spent fuel library

The SINRD and PDET techniques were investigated in this Ph.D. mainly through Monte Carlo simulations, so the development of a reliable model is of importance. A reference spent fuel library was developed in the first step of this Ph.D. research to obtain realistic material compositions and source terms for spent fuel with different irradiation histories. The neutron and gamma-ray emissions were defined in terms of source intensity and energy distribution.

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generate material composition and source terms compatible with the format of the MCNPX code that was used to investigate the NDA techniques. In a broader scope, the data from the fuel library are publicly available and can be used for the development of other NDA methods or for studies on the final disposal of spent fuel.

Monte Carlo study on SINRD

The passive neutron emission from spent fuel is measured with SINRD, and the attenuation of the neutron flux around the 0.3 eV energy region is used to directly quantify the 239Pu mass. The microscopic cross-section of 239Pu shows a strong resonance around 0.3 eV. The cross-section expresses the interaction probability between a certain nuclide and an incoming neutron, and it is specific for each nuclide. Therefore, significant neutron absorption is expected in correspondence of the 0.3 eV resonance due to the presence of 239Pu in spent fuel.

A thin foil of either Gd or Cd is placed around the neutron detector during the SINRD measurements. These elements were chosen because they show a cutoff energy for neutron absorption slightly below and above 0.3 eV, respectively. Due to this property these materials are called SINRD filters. By taking the difference of the neutron counts measured with the two filters, the neutron flux in the energy region close to the 239Pu resonance is estimated.

The Monte Carlo modelling of the SINRD technique was used to identify the optimal measurement setup. The approach proposed in this Ph.D. foresees the introduction of small neutron detectors in the central guide tube of the PWR 17x17 fuel assembly.

The measurement of fuel assembly immersed in fresh and borated water was modelled, and compared to the case of fuel kept in air and surrounded by a thick slab of polyethylene. The results from the dry configuration showed the clearest reduction of the neutron flux due to the absorption of 239Pu. The dry configuration was chosen as reference condition for the study and can be representative of a measurement station in an encapsulation plant for the final verification of a fuel assembly before the insertion in the storage canister for geological disposal.

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The detector response of a 239Pu fission chamber was compared with the results obtained with a 235U fission chamber and with proportional counters containing 3He or 10B. The 239Pu fission chamber showed the highest sensitivity to the 239Pu content in the fuel and this is the advantage of using the self-indication technique. Similar results were calculated for the other detector types, with the proportional counters that obtained the highest total neutron counts thanks to an higher neutron sensitivity. Taking these results into account, a combination of SINRD filters of 0.1 mm Gd and 1.0 mm Cd was suggested for the measurements with fission chambers to maximize the total neutron counts, whereas SINRD filters of 0.2 mm Gd and 1.0 mm Cd were proposed for the proportional counters to maximize the SINRD signature.

The expected performance of SINRD in realistic scenarios was evaluated by considering a detailed fuel composition. The masses of 239_{Pu and}235_{U were the parameters that influenced the most the} technique, while a few other nuclides had an impact in case of fuel with high burnup. The SINRD signature increased with the burnup due to the 239Pu content, and with the initial enrichment due to the 235U mass. Moreover, the SINRD signature was largely independent from the cooling time of the fuel assembly, since the fissile content in the fuel does not depend on this parameter. The approach proposed in this study showed also no significant effect from the positioning of detector in the guide tube, detector length, and small variations from the nominal filter thickness. The incomplete detector cover by the filters caused a change in the SINRD signature values due to the increase of the thermal neutron components passing through the bare section of the detector.

SINRD experimental benchmark

The results from the Monte Carlo study of SINRD were supported by an experimental benchmark carried out at the GELINA Time-of-Flight (ToF) facility of the Joint Research Centre (JRC) of Geel (Belgium). The Time-of-Flight technique was chosen for the experimental validation of SINRD because with this technique the energy distribution of a neutron beam can be measured. Time-of-Flight measurements are traditionally used for neutron resonance spectroscopy and measure the time that a neutron needs to travel a given distance. The measured time and the flight distance are then related to the kinetic energy of the neutron.

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In addition, self-indication measurements were carried out to confirm the basic principle of SINRD. Measurements on natural Cd samples with different thicknesses were used to mimic the presence of 239

Pu in spent fuel. To obtain the self-indication technique a thin Cd sample was surrounded by 4 C6D6 liquid scintillators detecting the prompt -rays emitted after (n,) reactions. Measurements were performed with Gd and Cd SINRD filters in the beam, and the thickness of these filters was optimized for the detection of neutrons with energy close to the Cd resonance at 0.178 eV. The results obtained with the self-indication geometry were compared with measurements with Frisch gridded ionization chambers with thin deposits of 235U or 10B. The self-indication detector showed an enhanced efficiency at the energy of the resonance of interest, i.e. the 0.178 eV resonance.

The results obtained for the self-indication experiments using the SINRD filters were very similar to the values calculated with the ideal measurement using ideal filters and background subtraction. The comparison of the results from the self-indication measurements with the values obtained for the other detectors confirmed that the highest sensitivity is obtained using a neutron detector with an enhanced efficiency for a resonance of the material of interest. Therefore, a 239Pu fission chamber is recommended for the characterization of spent fuel by SINRD.

Monte Carlo study on PDET

The partial defect tester (PDET) consists of a set of neutron and gamma detectors to measure the spontaneous emission from spent fuel. Several small detectors are simultaneously inserted from the top in the guide tubes of a PWR fuel assembly. These locations are designed for the insertion of the control rods when the assembly is loaded in the reactor core, and they are generally empty once the assembly is stored in the spent fuel pool. The measurement is performed without moving the fuel assembly from the storage location. This Ph.D. work considered 235U fission chambers and 238U fission chambers for the detection of thermal and fast neutrons, respectively. In addition, ionization chambers are used for the measurement of the gamma flux. The PDET detector was conceived for the partial defect verification of spent fuel, as the removal of fuel pins alters the spatial distribution of the neutron and gamma fluxes across the fuel assembly cross-section and allows the detection of the diversion.

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The detector responses of the 235U fission chambers were compared with the results from 239Pu fission chambers and proportional counters containing 3He and 10B, but no significant difference was observed. Ionization chambers containing N and Xe as filling gas at different pressures were compared as well, but the difference in the responses were within the statistical uncertainty.

The initial enrichment and cooling time of the spent fuel assembly did not affect significantly the normalized detector responses across the fuel assembly cross-section, whereas the fuel burnup had an effect within ±3% on the normalized neutron detector responses.

The influence of the irradiation history of the fuel assemblies in the storage rack on the normalized detector responses calculated in the guide tubes of the central fuel assembly was evaluated. The burnup of the central fuel assembly had the largest impact in case of storage racks with high burnup fuel assemblies in the lateral and corner positions. The neutron detectors in the guide tubes at the periphery of the assembly were mostly affected by the change in burnup of the central fuel assembly, and differences within ±10% were calculated on the normalized detector responses compared to the reference case.

Several storage rack configurations were also developed to estimate the influence of the lateral and corner fuel assemblies, and for all cases the maximum differences on the normalized detector responses compared to the reference case were obtained in the guide tubes close to the assemblies with high burnup. Variations between -50% and +30% were calculated for different guide tubes in the central fuel assembly. Because of the self-shielding effect of the fuel pins, the gamma-ray detector responses were in general less influenced by the fuel assemblies with different burnup compared to the responses of the fission chambers.

Comparison of the partial defect capabilities for SINRD and PDET

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For both techniques the insertion of multiple detectors in the different guide tube positions was modelled and the detector responses were normalized to the maximum value obtained among the guide tubes. Based on the Monte Carlo studies the detector responses of bare 238U fission chambers as well as 239Pu fission chambers covered by SINRD filters were considered for SINRD, whereas the responses of 235U fission chambers, 238U fission chambers, and ionization chambers were calculated for PDET.

Significant differences in the normalized detector responses were observed for both SINRD and PDET in the diversion scenarios with 50% of fuel pins replaced, and the most challenging scenario was obtained with the replacement with a chess-board pattern.

In the case of SINRD the detector responses of the 238U fission chambers were more affected by the fuel pins diversion than the 239Pu fission chambers covered by the SINRD filters. For both detector types the average detector response showed a difference between -30% and +15% in the case of 50% dummy pins.

The results for the PDET detector showed that the detector responses of the gamma-ray detectors have the highest sensitivity to the diversion scenarios. As for SINRD, the variation in the detector responses among the guide tubes provides indications for the detection of diversion scenarios with 50% of dummy pins. Relative differences up to -30% from the reference case were calculated for the ionization chambers in the guide tubes at the periphery of the assembly. The variations for the neutron detectors were within ±20% for the peripheral guide tubes and within ±10% for the central guide tubes.

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Résumé

Introduction

Ce projet de doctorat décrit le développement de méthodes de contrôle non-destructif (CND) de mesure du combustible nucléaire irradié et a été mené en collaboration entre l'Université Libre de Bruxelles (ULB) et le centre de recherche nucléaire belge SCK•CEN.

Le combustible nucléaire irradié est l’ensemble des assemblages de combustible qui sont déchargés des réacteurs nucléaires après irradiation et qui sont transférés dans un stockage intermédiaire. Ce combustible contient des éléments radioactifs qui sont responsables des émissions de neutrons et de rayons gamma. Puisque la désintégration radioactive produit également une quantité de chaleur significative, le combustible irradié est normalement stocké sous eau pour assurer un refroidissement approprié et pour fournir une protection contre les rayonnements.

Après l'irradiation, le combustible irradié contient encore environ 2% de matières fissiles (235U et 239

Pu). Par conséquent, les mécanismes de protection nucléaire sont d’application afin de garantir l’utilisation du matériau à des applications uniquement pacifiques. Le plutonium contenu dans le combustible irradié est une préoccupation majeure pour la communauté de protection contre la prolifération, car il représente près de 80% de toutes les matières placées sous protection aujourd'hui. De plus, la décharge du combustible irradié des réacteurs en exploitation conduit à l'augmentation cumulative de l'inventaire au fil du temps.

Plusieurs techniques de CND sont utilisées pour vérifier le combustible irradié et des techniques supplémentaires sont en cours de développement pour fournir des mesures plus précises. Des techniques passives et actives sont considérées. Les techniques passives reposent sur l'émission spontanée de rayonnements provenant du combustible irradié, alors que des sources externes sont utilisées pour des techniques actives. Le rayonnement mesurée est différent pour chaque méthode CND, il est possible d’utiliser les neutrons, les rayons gamma ou le rayonnement Cherenkov. Les techniques étudiées dans cette thèse de doctorat sont le Self-Indication Neutron Resonance Densitometry (SINRD) et le Partial Defect Tester (PDET).

Développement de la bibliothèque du combustible irradié

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sources pour le combustible irradié avec différentes histoires d'irradiation. Les émissions de neutrons et de rayons gamma ont été définis en termes d'intensité de la source et de distribution d'énergie. Le développement de la bibliothèque de combustible a permis de comprendre l'influence de l'irradiation du combustible sur la composition isotopique et par conséquent sur l’intensité de la source neutronique et de rayonnement gamma provenant du combustible irradié. En outre, une comparaison entre ORIGEN-ARP and ALEPH-2.2 utilisés pour le calcul a également été réalisée. L'objectif final dans le cadre du doctorat était de générer composition du matériel et termes de source compatibles avec le code MCNPX qui a été utilisé pour étudier les techniques CND. Dans un cadre plus large, les données de la bibliothèque de combustible sont accessibles au public et peuvent être utilisées pour le développement d'autres méthodes CND ou pour des études sur le stockage définitif du combustible irradié.

Étude Monte Carlo de la méthode SINRD

L'émission passive de neutrons du combustible irradié est mesurée avec SINRD, et l'atténuation du flux de neutrons autour de la région d'énergie de 0.3 eV est utilisée pour quantifier directement la masse de 239Pu. La section efficace microscopique de 239Pu montre une forte résonance autour de 0.3 eV. La section efficace exprime la probabilité d'interaction entre un certain nucléide et un neutron entrant, et il est spécifique pour chaque radionucléide. Par conséquent, une absorption importante de neutrons est prévisible en correspondance de la résonance à 0.3 eV dû à la présence de 239Pu dans le combustible irradié.

Une mince feuille soit de Gd ou de Cd est placée autour du détecteur de neutrons pendant les mesures de SINRD. Ces éléments ont été choisis parce qu'ils montrent une énergie de coupure pour l’absorption de neutrons légèrement au-dessous et au-dessus de 0.3 eV. En raison de cette propriété, ces matériaux sont appelés filtres SINRD. En prenant la différence entre les comptages de neutrons mesurés avec les deux filtres, le flux de neutrons dans la région de l'énergie proche de la résonance 239

Pu est estimé.

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référence pour l'étude et peut être représentative d'une station de mesure dans une usine d'encapsulation pour la vérification finale d'un assemblage combustible avant l'insertion dans le container de stockage pour le stockage géologique.

La signature SINRD a été définie comme étant le rapport entre les comptages de neutrons dans la zone rapide et dans la région d'énergie de résonance à 0.3 eV. Une chambre à fission 238U a été choisie comme détecteur de référence pour l'estimation du flux de neutrons rapides, alors qu’une chambre à fission 239Pu recouverte par une feuille soit de Gd ou de Cd a été proposée pour la région de résonance. L'optimisation de l'épaisseur du filtre SINRD a également été réalisée, en identifiant une combinaison de filtres conduisant principalement à des contributions de neutrons avec une énergie proche de 0.3 eV et à minimiser les contributions provenant d'autres zones d'énergie.

La réponse du détecteur d'une chambre à fission de 239Pu a été comparée aux résultats obtenus avec une chambre à fission de 235U et des compteurs proportionnels contenant du 3He ou du 10B. La chambre à fission de 239Pu a montré la plus grande sensibilité à la teneur en 239Pu dans le combustible, ce qui est l'avantage d'utiliser la technique de "self-indication". Des résultats similaires ont été calculés pour les autres types de détecteurs, avec les compteurs proportionnels qui ont obtenu le plus haut total de comptages de neutrons grâce à leur sensibilité élevée aux neutrons. Compte tenu de ces résultats, une combinaison de filtres de SINRD de 0.1 mm de Gd et 1.0 mm de Cd a été suggérée pour les mesures avec des chambres à fission pour maximiser le total des comptages de neutrons, alors que les filtres SINRD de 0.2 mm de Gd et 1.0 mm de Cd ont été proposés pour les compteurs proportionnels pour maximiser la signature SINRD.

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Référence expérimentale SINRD

Les résultats de l'étude Monte Carlo de la méthode SINRD ont été soutenus par une analyse expérimentale réalisée au GELINA Time-of-Flight (ToF), une infrastructure du Joint Research Centre (JRC) de Geel (Belgique). La technique Time-of-Flight a été choisie pour la validation expérimentale de la méthode SINRD parce qu'avec cette technique, la répartition d'énergie d'un faisceau de neutrons peut être mesurée. Des mesures Time-of-Flight sont traditionnellement utilisées pour la spectroscopie par résonance neutronique et mesurer le temps dont un neutron a besoin pour parcourir une distance connue. Le temps mesuré et la distance de vol sont ensuite liés à l'énergie cinétique du neutron.

Des mesures de transmission ont été effectuées pour vérifier la qualité des données nucléaires utilisées dans les simulations pour l'optimisation des filtres SINRD. La comparaison a montré quelques différences entre la transmission expérimentale et les valeurs calculées par l'approche analytique. Cependant, la qualité des données nucléaires est suffisante pour définir l'épaisseur optimale des filtres Gd et Cd. Les résultats des expériences ont indiqué que la combinaison d'un filtre de Gd environ 0.1 mm d'épaisseur avec un filtre Cd de 1.0 mm est appropriée pour la mesure du combustible irradié contenant du 239Pu.

En outre, les mesures de self-indication ont été effectuées pour confirmer le principe de base de la SINRD. Les mesures sur des échantillons de Cd naturel avec des épaisseurs différentes ont été utilisées pour simuler la présence de 239Pu dans le combustible irradié. Pour construire un détecteur avec une sensibilité élevée dans la région d'énergie proche de la résonance à 0.178 eV, un échantillon de Cd mince a été entouré par 4 scintillateurs liquides C6D6 détectant les rayons  prompts émis par réactions (n, ). Pour améliorer la sensibilité aux alentours de 0.178 eV, les mesures ont été effectuées avec des filtres SINRD en Gd et Cd interposés dans le faisceau. Les résultats obtenus avec la géométrie de self-indication ont été comparées aux mesures effectuées avec les chambres d'ionisation à grille de Frisch avec de minces dépôts de 235U ou 10B. Le détecteur de self-indication présente une efficacité accrue à l'énergie de la résonance d'intérêt, à savoir la résonance à 0.178 eV.

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chambre à fission 239Pu est recommandée pour la caractérisation du combustible irradié par la méthode SINRD.

Étude Monte Carlo sur PDET

Le Partial Defect Tester (PDET) consiste en un ensemble de détecteurs de neutrons et gamma pour mesurer l'émission spontanée du combustible irradié. Plusieurs petits détecteurs sont insérés simultanément par le haut dans les tubes-guides d'un assemblage combustible de PWR. Ces emplacements sont conçus pour l'insertion des barres de contrôle, lorsque l'ensemble est chargé dans le cœur du réacteur, et ils sont généralement vides une fois que l'assemblage est stocké dans la piscine du combustible irradié. La mesure est effectuée sans déplacer l'assemblage de combustible à partir de son emplacement de stockage. L’étude effectuée dans ce doctorat a considéré des chambres à fission en 235U et 238U pour la détection des neutrons thermiques et rapides, respectivement. En outre, des chambres d'ionisation sont utilisées pour la mesure du flux de rayons gamma. Le détecteur PDET a été conçu pour la vérification de défaut partiel du combustible irradié, comme le retrait des crayons combustibles modifie la distribution spatiale des flux à travers la section transversale de l'assemblage combustible et permettre la détection de la déviation.

Le modèle du rack de stockage est composé de neuf assemblages combustibles dans une configuration 3x3 avec le PDET inséré dans l'assemblage combustible central. La fonction d'importance de chaque crayon combustible de l'assemblage combustible mesuré a été calculée par simulations Monte Carlo pour les différents détecteurs de neutrons et de rayons gamma. Tous les crayons combustibles ont contribué d'une manière significative aux réponses des deux types de chambre de fission, tandis que les contributions aux chambres d'ionisation sont fortement localisées dans le voisinage du tube-guide contenant le détecteur.

Les réponses des détecteurs des chambres à fission de 235U ont été comparées avec les résultats des chambres de fission de 239Pu et des compteurs proportionnels contenant du 3He et 10B, mais aucune différence significative n'a été observée. Les chambres d'ionisation contenant du N et Xe comme gaz de remplissage à des pressions différentes ont été comparées de la même manière, mais la différence dans les réponses se situe dans l'incertitude statistique.

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L'influence de l'irradiation des assemblages combustibles dans le rack de stockage sur les réponses normalisées des détecteurs calculées dans les tubes-guides de l'assemblage combustible central a été évaluée. Le burnup de l'assemblage combustible central a le plus grand impact dans le cas des supports de stockage avec des assemblages combustibles à haut burnup dans les positions latérales et de coin. Les détecteurs de neutrons dans les tubes-guides à la périphérie de l'assemblage ont été principalement affectés par le changement de burnup de l'assemblage combustible central, et des différences dans un intervalle de ± 10% ont été calculées sur les réponses normalisées des détecteurs par rapport au cas de référence.

Plusieurs configurations de rack de stockage ont également été développées pour estimer l'influence des assemblages de combustible latérales et de coin, et pour tous les cas, les différences maximales sur les réponses normalisées des détecteurs par rapport au cas de référence ont été obtenues dans les tubes-guides à proximité des assemblages avec des burnups élevés. Des variations entre -50% et + 30% ont été calculées pour les différents tubes-guides de l'assemblage combustible central. En raison de l'effet d'auto-protection des barres de combustible, les réponses des détecteurs de rayons gamma sont en général moins influencées par les assemblages de combustible avec un burnup différent par rapport aux réponses des chambres à fission.

Comparaison des capacités de défauts partiels pour SINRD et PDET

Les détecteurs SINRD et PDET ont été comparés aux simulations Monte Carlo quant à leur capacité à détecter les crayons combustibles remplacés par des crayons factices. L'objectif actuel de l'AIEA pour les tests de défaut partiel est de vérifier qu’au moins 50% des crayons combustibles sont présents dans un assemblage combustible. Par conséquent, une série de 12 scénarios de modification ont été créés pour cette analyse, en tenant compte de remplacements de 50% à 15% des crayons combustibles de l'assemblage combustible mesuré. Les crayons combustibles détournés ont été remplacés par des substituts en acier inoxydable avec les mêmes dimensions que les crayons combustibles irradiés d'origine. Un motif symétrique a été choisi pour tous les cas, car il a été considéré comme la situation la plus difficile à détecter. Une modification uniforme a été prise en compte, ainsi que les cas de remplacement des crayons dans la section extérieure et de la section intérieure de l'assemblage.

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SINRD ont été considérées pour SINRD, alors que les réponses des chambres à fission 235U, chambres à fission238U, et les chambres d'ionisation ont été calculées pour PDET.

Des différences significatives dans les réponses normalisées des détecteurs ont été observées pour SINRD et PDET dans les scénarios de dérivation avec 50% des crayons combustibles remplacés, et le scénario le plus délicat a été obtenu avec le remplacement d'un motif en échiquier.

Dans le cas de SINRD les réponses des détecteurs des chambres à fission 238U ont été plus affectées par le détournement des crayons combustibles que les chambres à fission 239Pu couverts par les filtres SINRD. Pour les deux types de détecteurs des différences relatives comprises entre -30% et +15% ont été calculées à partir du cas de référence compte tenu de ces deux types de détecteurs. Les résultats obtenus pour le détecteur de PDET a montré que les réponses des détecteurs de rayons gamma ont la plus grande sensibilité aux scénarios de modification. En ce qui concerne SINRD, la variation dans les réponses moyennes des détecteurs et les écarts maximaux entre les tubes-guides fournissent des indications pour la détection des scénarios de dérivation avec 50% de crayons factices. Des différences relatives jusqu'à -30% par rapport au scénario de référence ont été calculées pour les chambres d'ionisation dans les tubes-guides à la périphérie de l'assemblage. Les variations pour les détecteurs de neutrons sont à moins de ±20% pour les tubes-guides périphériques et à ±10% pour les tubes-guides central.

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1 Introduction

1.1 The framework of nuclear safeguards

After the discovery of nuclear fission in 1939, nuclear energy was used both for civil and military purposes. On the one hand the first man-made self-sustaining nuclear chain reaction was achieved in December 1942 with the Chicago Pile 1, and the first nuclear power plants were connected to the power grid during the 1950s. On the other hand the first nuclear weapon was tested in July 1945 at the Trinity Site in the USA, followed by the bombings of Hiroshima and Nagasaki in August 1945 (Rhodes, 1986).

The risk of the proliferation of nuclear weapons after the Second World War led to the foundation of the International Atomic Energy Agency (IAEA) in 1957. According to the IAEA statute (IAEA, 1989), the Agency shall support the contribution of nuclear energy to world-wide peace, health, and prosperity. Moreover, it shall ensure that nuclear technology and knowledge is not used to pursue military purposes.

The Article III.5 of the IAEA statute authorizes the Agency "to establish and administer safeguards

designed to ensure that special fissionable and other materials, services, equipment, facilities, and information ... are not used in such a way to further any military purpose" (IAEA, 1989). The legal

basis was granted by the treaty on the non-proliferation of nuclear weapons (NPT) (IAEA, 1970). Five countries signatories of the NPT – United States, Russian Federation, United Kingdom, France, China – are defined nuclear weapon states (NWS) since they detonated nuclear weapons before the entry into force of the treaty, whereas all other countries are considered non-nuclear weapon states. By signing the NPT all parties agree to enhance collaboration on the peaceful uses of nuclear energy, and accept safeguards measures to prevent the dissemination of nuclear weapons. Moreover, the NWS agree to reduce the existing stockpiles and to reach nuclear disarmament.

The technical objective of safeguards was defined in the INFCIRC/153 (IAEA, 1972) as "the timely

detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection."

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As a response to these findings the IAEA drafted the Additional Protocol (AP) (IAEA, 1997), which supports the NPT and focuses on the detection of clandestine military programs. The AP grants more access rights to the inspectors and the state declaration covers a wider range of facilities and activities compared to the NPT verification regime based on the INFCIRC/153.

The two agreements are complementary to each other; the NPT ensures the absence of diversion of declared nuclear material and verifies the declared use of facilities, whereas the AP ensures the absence of undeclared nuclear material and activities.

Focusing on the European scale, the EURATOM treaty (European Union, 2010) states that the European Commission shall ensure that "ores, source materials and special fissile materials are not

diverted from their intended uses as declared by the users". To satisfy this requirement, the

EURATOM safeguards system was established as a set of controls and verifications covering the nuclear fuel cycle from mining to final storage. The provisions for supply and safeguards obligations must be satisfied also for agreements with non-member states and international organizations (European Union, 2010). EURATOM safeguards, as part of the European Commission’s Directorate-General for Energy (DG-ENER), serve also as focal point in the EU member states for the implementation of the safeguards agreement and additional protocol with the IAEA (European Commission, 2014).

1.2 Renewed interest for spent fuel measurement methods

Among the safeguards verification activities, the measurement of spent fuel was identified by EURATOM safeguards as one of the major challenges. Thus significant R&D activities are carried out within the European Commission and through collaborations with EU and non-EU countries and the IAEA (Schwalbach, 2014). The main projects deal with the validation of a software to compare the expected neutron and gamma count rates and the measurements with a Fork detector (Vaccaro, 2014). Moreover, the development of the partial defect tester and the gamma-emission tomography are carried out (Goncalves, 2014; Tobin, 2014).

Similarly to EURATOM safeguards, the IAEA department of safeguards long term R&D plan for 2012-2023 identified as topic with high priority the development of NDA methods for the spent fuel verification before the transfer to difficult-to-access storage (IAEA, 2013).

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The Swedish and Finnish support programs to the IAEA focused on the development of gamma-ray emission tomography and Monte Carlo simulations and experimental tests gave encouraging results (Honkamaa, 2014; Jansson, 2015). With this technique a series of medium-resolution gamma-ray detectors rotates around a spent fuel assembly and by image reconstruction techniques a tomographic image of the fuel pins is obtained. The instrument is dedicated to the partial defect verification and burnup estimation of spent fuel.

The Next Generation Safeguards Initiative – Spent Fuel (NGSI-SF) is a joint project between several universities and US national laboratories (Tobin, 2011). The NGSI project studied 14 NDA techniques and after an external evaluation focused on four combinations of single techniques for further study (Charlton, 2012). Both active and passive techniques were considered, covering both neutron and gamma detectors. The integration with current NDA methods is foreseen in the design of the prototypes.

1.3 Structure and objectives of the research project

This Ph.D. project aimed at investigating the Self-Indication Neutron Resonance Densitometry (SINRD) and the Partial Defect Tester (PDET). These are two NDA techniques proposed for the measurement of spent nuclear fuel. The research was based mainly on Monte Carlo simulations, and a reference spent fuel library was developed as preparatory step to obtain a reliable composition and source term of the spent fuel assembly.

An overview of the safeguards challenges for spent fuel measurements is given in Chapter 2. This section includes also a description of the current NDA techniques used for spent fuel verifications, as well as the two techniques studied in this Ph.D. topic.

The approach chosen in the study is outlined in Chapter 3, in addition to the overview of previous work on SINRD and PDET. The Monte Carlo models are described in terms of fuel assembly geometry and storage configurations. Chapter 3 includes also the explanation on the contribution of the reference spent fuel library to this research project, and the approach developed for the neutron and gamma-ray detectors responses concludes the section.

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Finally the performance in realistic scenarios were modeled by considering several spent fuel compositions and a sensitivity study on the SINRD technique.

The experimental benchmark of SINRD is described in Chapter 5. The measurements were carried out at the GELINA Time-of-Flight facility at JRC-IRMM in Geel (Belgium). The goals of the benchmark were to verify the quality of the nuclear data used in the Monte Carlo simulations and to perform self-indication experiments on a set of target samples. Moreover, the responses of several detector types were compared.

The results from Monte Carlo study of the PDET detector are given in Chapter 6. Reference conditions were defined first to calculate the contribution of single fuel pins on the detector response. The responses of several detector types were compared through simulations. Then the influence of the different fuel assemblies inserted in the storage rack was evaluated by considering assemblies with different compositions and source terms.

The performance of SINRD and PDET for the partial defect detection is compared in Chapter 7. Several diversion scenarios were developed with the Monte Carlo model by removing fuel pins from complete fuel assemblies.

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2 Safeguards challenges of spent nuclear fuel

2.1 Properties of spent nuclear fuel

The term of spent nuclear fuel refers to fuel assemblies discharged from nuclear reactors after irradiation (IAEA, 2012a). The irradiation of fresh fuel in the reactor core leads to the production of several fission products and minor actinides, some of which are strong neutron absorbers. In addition, the total fissile material content decreases due to the fuel burnup, and therefore the fuel assemblies are unloaded from the core after a few irradiation cycles and transferred to an interim storage.

Spent fuel contains many radioactive elements that are responsible for very strong neutron and gamma emissions. Since the radioactive decay is responsible also for significant decay heat, the spent fuel is normally stored under water to ensure appropriate cooling and provide radiation shielding. The average composition of spent Low Enriched Uranium (LEU) fuel from Light Water Reactors (LWR) with 3.5% initial enrichment, 33 GWd/tHM discharge burnup, and 3 years cooling time is shown in Figure 2-1. About 95% of spent fuel consists of 238U, whereas slightly more than 3% are fission products and minor actinides. The fissile materials, namely 235U and Pu, are almost in the same quantity and combined account for 2% of the total mass (IAEA, 2012b).

Figure 2-1: Material composition of LEU spent fuel with 3.5% initial enrichment, 33 GWd/tHM burnup, and 3 years cooling

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Although it accounts only for a fraction of the total spent fuel, the residual fissile material is a main concern for the safeguards community. The plutonium contained in spent fuel and in fuel assemblies in reactor cores represents almost 80% of the material placed under safeguards today (IAEA, 2014). According to (IAEA, 2012b) more than 11000 tHM of spent fuel are discharged every year worldwide and the total amount of spent fuel will reach approximately 450000 tHM by 2020. Figure 2-2 shows the trend of the cumulative amount of spent fuel, taking into account storage and reprocessing.

Figure 2-2: Cumulative inventory of spent fuel generated worldwide. (IAEA, 2012b)

2.2 Safeguards requirements for spent fuel verifications

The spent fuel verifications are carried out through nuclear material accountancy (NMA) and containment and surveillance (C/S) measures, according to the principles outlined in INFCIRC/153 (IAEA, 1972).

The NMA is performed by counting the number of spent fuel assemblies and by NDA measurements to detect the diversion of nuclear material. These activities are normally carried out during the annual physical inventory verification (IAEA, 1998). The NDA techniques are used for the gross and partial defect testing. The gross defect testing aims at differentiating spent fuel assemblies from other materials such as fresh fuel or structural materials. The IAEA safeguards criteria (IAEA, 2009) defines that the partial defect testing "should ensure that at least half of the fuel pins are present in

each fuel assembly".

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material in the surveillance zone. The records are then compared with the declared inventory change to confirm the operator declaration. (IAEA, 1998)

2.3 Current non-destructive assays for spent fuel measurements

2.3.1 Overview of NDA techniques

Several NDA techniques were developed in the past for spent fuel measurements, and other methods are investigated to improve the current capabilities.

A list of such NDA techniques is included in Table 2-1, and those techniques span from in-field use to research projects. Both passive and active techniques are considered. Passive techniques rely on the spontaneous emission of radiation from the spent fuel itself, whereas external radiation sources are used for active techniques. The radiation used by each NDA method is also specified, and varies between neutron, gamma-ray, and Cherenkov light. No radiation is mentioned for calorimetry, since the technique measures the increase of temperature of the medium surrounding the fuel assembly due to the residual decay heat. Finally, the proposed goals of each technique are mentioned with known issues. The known issues and limitations of the techniques will be discussed in detail in the next chapters.

The first part of the table includes the NDA techniques used currently in-field, and more information on each method is provided in the next sections. The SINRD and PDET techniques investigated in this Ph.D. project are placed in the second part of the table and are described in details in the other chapters of this thesis. The last section of the table includes a list of other NDA methods under research, and detailed information on each technique can be found in (Honkamaa, 2014; Jansson, 2015; Tobin, 2013; Schillebeeckx, 2014).

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Table 2-1: Overview of NDA techniques for spent fuel measurements. The acronyms of the techniques are included in the

list of acronyms at the beginning of the thesis.

Technique Status Type Radiation Proposed goals Known issues

DCVD Field use Passive Cherenkov Gross defect; Detector alignment; measurement of long-cooled fuel

partial defect

SFAT Field use Passive Gamma-ray Gross defect Self-shielding effect of fuel pins

Fork Field use Passive Neutron; gamma-ray

Gross defect Need for calibration curve

SINRD Research Passive Neutron 239Pu quantification; Detector availability; partial defect measurements in air PDET Prototype Passive Neutron;

gamma-ray

Partial defect; Intrusive;

burnup estimation detector calibration

Calorimetry Prototype Passive --- Residual heat Long measurement time

UGET Prototype Passive Gamma-ray Partial defect Indirect measurement

DDA Prototype Active Neutron Multiplication; Need for external neutron source verify operator data

DDSI Prototype Passive Neutron Multiplication Indirect measurement

NRD Demonstr. Active Neutron; gamma-ray

Isotopic quantification

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The Cherenkov radiation is emitted when a charged particle travels in a medium with a speed higher than the speed of light in that medium (Phillips, 1991).

Cherenkov radiation in spent fuel is due to high-energy gamma-rays emitted from the radioactive decay of fission and activation products. The interactions between these gamma rays and fuel cladding or storage water produce electrons and positrons that have sufficient energy to emit Cherenkov radiation (Phillips, 1991).

The Improved Cherenkov Viewing Device (ICVD) and the Digital Cherenkov Viewing Device (DCVD) are among the main NDA instruments used by IAEA for gross and partial defect testing (Chen, 2009; IAEA, 2011; Branger, 2014). These instruments detect the Cherenkov radiation induced when spent fuel is stored under water in a spent fuel pool, and are shown in Figure 2-3.

Figure 2-3: Improved Cherenkov viewing device (ICVD, left, (IAEA, 2011)). Digital Cherenkov viewing device (DCVD, right,

photo by Andy Gerwing, Channel Systems).

Both detectors intensify the ultraviolet light associated with the Cherenkov radiation and filter most of the visible light coming from the measurement environment. The ICVD is a hand-held device where one inspector can see the filtered image, whereas the DCVD is equipped with an external monitor allowing the complete inspection team to view the image.

The measurements with both ICVD and DCVD are performed on the crane over the spent fuel pool and therefore fuel movement or insertion of components in the storage pool is not required.

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Figure 2-4: Color-enhanced image obtained with the DCVD detector of BWR 8x8 (left, adapted from (Chen, 2003)) and a

PWR 17x17 fuel assembly (right, (Chen, 2009)).

The measurements of Cherenkov radiation with ICVD and DCVD are relatively fast and do not require the movement of fuel assemblies. On the other hand, the correct alignment between the detector and the fuel assembly plays an important role in the data analysis. The detection of missing fuel pins is challenging when the pins are substituted at the periphery of the assembly, or under the handle of the BWR fuel assembly (Parcey, 2011).

Current research is carried out on the measurement of long-cooled fuel when this is stored close to relatively short-cooled assemblies (Branger, 2014).

2.3.3 Spent fuel attribute tester

The Spent Fuel Attribute Tester (SFAT) measures the gamma-ray energy spectra from spent fuel and it is used for the gross defect verification (IAEA, 2011).

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Figure 2-5: Spent fuel attribute tester (SFAT) and water-tight collimator. (Chichester, 2009)

The presence of spent fuel is verified with SFAT by measuring the main gamma emission from 137Cs at 662 keV. Moreover, the measurement of the gamma-ray line from 144Pr at 2182 keV is possible for fuel assemblies with cooling time shorter than 4 years, as well as emissions from 134Cs, 154Eu, and 60Co (Janssen-Maenhout, 2008).

To confirm the presence of spent fuel, the intensity of the detected gamma-rays is compared with measurements with the detector placed in the gap between neighboring fuel assemblies.

As for the Cherenkov viewing devices, the measurements with SFAT do not require the movement of spent fuel. The use of SFAT is of interest when the measurement of Cherenkov radiation is difficult, for instance when measuring fuel with low burnup and long cooling time, or when the water is not clear enough (IAEA, 2011).

However, the SFAT gives information only from the top part of the fuel assembly, since the gamma-rays emitted from the lowest section of the fuel are shielded by the fuel assembly itself. This limitation rules out the SFAT for the verification of assemblies with partial length rods (Arit, 1995).

2.3.4 Fork detector

The Fork detector provides a measurement of the neutron and gamma emissions from a spent fuel assembly. Most of the spontaneous neutrons are emitted by 242Cm and 244Cm, whereas 137Cs is the main gamma emitter for fuel with a few years of cooling time. Several detector designs were developed and are currently used in several locations by IAEA and DG-ENER inspectors (Rinard, 1988).

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neutrons and gamma-rays detection, respectively (Rinard, 1984). One of the fission chambers is wrapped by a cadmium foil in order to be sensitive mainly to epithermal neutrons, whereas the other fission chamber is bare and is sensitive mainly to thermal neutrons.

Another version of the Fork detector was developed by SCK•CEN (Carchon, 1987; Carchon, 1994), and contains only one fission chamber and one ionization chamber in each detector arm. In this design, the polyethylene block is surrounded by a cadmium sheet to absorb thermal neutrons and a stainless steel housing (Borella, 2010). Both versions are shown in Figure 2-6.

Figure 2-6: Fork detector developed by LANL (left, (Antech, 2015)). Fork detector developed by SCK•CEN (right, (Borella,

2011)).

The Fork detector is introduced in the spent fuel pool and the measured fuel assembly is lifted partially from the storage rack position. The arms of the detector surround the fuel assembly and the neutron and gamma counts are collected simultaneously. In general, only one measurement close to the central axial region of the assembly is required for the verification (IAEA, 2011).

The measurement of neutron and gamma emissions is used for the gross defect testing, whereas the neutron signals coming from the bare and Cd-covered fission chambers give information on the boron content in the spent fuel pool (Phillips, 1991). The ratio between the neutron and gamma signals gives information about the fuel assembly burnup and cooling time; however, calibration curves must be established to achieve reliable burnup estimation (Rinard, 1986; Rinard, 1988; Borella, 2011).

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2.4 Techniques investigated in this Ph.D.

2.4.1 Self-indication neutron resonance densitometry

The self-indication neutron resonance densitometry (SINRD) measures the passive neutron emission from spent fuel (Menlove, 1969). The SINRD technique aims at measuring the attenuation of the neutron flux around the 0.3 eV energy region as a way to directly quantify the 239Pu mass in the spent fuel.

The microscopic cross-section of 239Pu is shown in Figure 2-7 and a clear peak, called resonance, around 0.3 eV is observed. The cross-section expresses the interaction probability between a certain nuclide and an incoming neutron, and it is specific for each nuclide. Therefore, significant neutron absorption is expected in correspondence of the 0.3 eV resonance due to the presence of 239Pu in spent fuel.

Figure 2-7: Total microscopic cross-section of 239Pu according to the ENDF-B/VII.0 nuclear data library (Chadwick, 2006). The right plot is a zoom on the energy region close to 0.3 eV.

A thin foil of either Gd or Cd is placed around the neutron detector during the SINRD measurements. These elements were chosen because they show a cutoff energy for neutron absorption slightly below and above 0.3 eV. Due to this property these materials are called SINRD filters. By taking the difference of the neutron counts measured with the two filters, the neutron flux in the energy region close to the 239Pu resonance is estimated.

2.4.2 Partial defect tester

The partial defect tester (PDET) consists in a set of neutron and gamma detectors to measure the spontaneous emission from spent fuel (Sitaraman, 2007). With the PDET detector several small detectors are simultaneously inserted from the top in the guide tubes of a PWR fuel assembly. These

10-9 10-7 10-5 10-3 10-1 101 100 101 102 103 104 105 T o ta l c ro s s -s e c ti o n ( b )

Neutron energy (MeV) 239 Pu 10-7 10-6 101 102 103 104 105 T o ta l c ro s s -s e c ti o n ( b )

Neutron energy (MeV) 239

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locations are designed for the insertion of the control rods when the assembly is loaded in the reactor core, and they are generally empty once the assembly is stored in the spent fuel pool.

A prototype of the detector was developed by Lawrence Livermore National Laboratory (LLNL) for the measurement of PWR 17x17 fuel assemblies, and Figure 2-8 shows a picture of the prototype. The neutrons are measured with 235U fission chambers, whereas ionization chambers are used for the measurement of the gamma flux.

The geometrical location of the guide tubes lead to a reference profile of the neutron and gamma fluxes across the fuel assembly cross-section. The removal of fuel pins from a fuel assembly alters the spatial distribution of the fluxes and should allow the detection of the missing fuel pins. Therefore, the PDET detector was conceived for the partial defect verification of spent fuel.

Advanced non-destructive methods for criticality safety and safeguards of used nuclear fuel