BOOK OF ABSTRACTS September 1-6, 2024 – Avignon, France 2 3 PLENARY TALKS ............................................................. 12 Global Overview on the Nuclear Fuel Cycle Backend and IAEA Related Activities ............................. 13 Clément Hill, Amparo Gonzalez-Espartero .................................................................................................................................................................................. 13 Status of the French Nuclear Fuel Cycle Program ................................................................................................... 15 François Sudreau............................................................................................................................................................................................................................................ 15 Nuclear Fuel Recycle Activities in the Office of Nuclear Energy ....................................................................... 17 Stephen Kung .....................................................................................................................................................................................................................................................17 Future Fuel Cycles – a UK Perspective ............................................................................................................................. 19 Paul Nevitt ............................................................................................................................................................................................................................................................ 19 ACTINIDE AND FISSION PRODUCTS SEPARATION ...... 22 Overview of the Material Recovery and Waste Form Development Program ..................................... 23 Kenneth C Marsden ..................................................................................................................................................................................................................................... 23 Lab-Scale Pulsed Columns Trials for a New Nuclear Fuel Recycling Process ...................................... 25 Garzon Losik German, Lamadie Fabrice, Roussel Hervé .................................................................................................................................................. 25 Potential of Aggregation Control for Solvent Extraction Separation .......................................................... 27 Cyril Micheau,a Yuki Ueda,a Ryuhei Motokawa,a Kazuhiro Akutsu-Suyama,b, c Norifumi L. Yamada,d Masako Yamada,d Sayed Ali Moussaoui,e Elizabeth Makombe,e Daniel Meyer,e Laurence Berthon,f Damien Bourgeoise ............................................. 27 Evolution of Uranium Recovery: Past, Present, and Future Perspectives .................................................. 29 Santa Jansone Popova1, Jopaul Mathew1, Jeffrey Einkauf1, Alexander Ivanov 1, Ilja Popovs1, Connor Parker1, Peter Zalupski2, Travis Grimes2, Corey Pilgrim2 ........................................................................................................................................................................................ 29 Experimental and Modeling Study of Uranium(VI) and Nitric Acid Extraction With a N,N- Dialkylamide Solvent ................................................................................................................................................................... 31 Thibau Blanc1, Donatien Gomes Rodrigues1, Pauline Moeyaert1*, Thomas Dumas1, Philippe Guilbaud1* ....................................... 31 Feasibility Study on PUREX-NUMAP Hybrid Reprocessing: Precipitation-Based Recovery of U(VI) from Organic Phases with 30% TBP ...................................................................................................................... 33 Ririka Tashiro1), Satoru Tsushima1)2), Koichiro Takao1).......................................................................................................................................................... 33 Demonstration of U(VI)/Pu(IV) Separation by Solvent Extraction in Modified Lab-Scale Annular Centrifugal Contactors Using D2EHiBA Extractant............................................................................... 35 Dominic Maertensa,b*, Koen Binnemansb, Thomas Cardinaelsa,b ............................................................................................................................. 35 Current TRL Status and Strategy for the Development of the Next Generation of Reprocessing Plant ....................................................................................................................................................................................................... 37 B. Arab-Chapelet1, C. Sorel1, I. Hablot2, A. Gil Martin2, M. Phelip3, A. Salvatores3, L. Diaz4, G. Vaast4 ..................................................... 37 Towards a Single-Solvent Process for U/TRU Recovery and Minor Actinide/Lanthanide Separations: Speciation and Partitioning of Tetravalent (Th, Pu) and Hexavalent (U) Actinides with HEH[EHP] and T2EHDGA .................................................................................................................................................. 39 Artem Gelis*, Joel Castillo, Logan Smith, Quinn Summerfield, Frederic Poineau .......................................................................................... 39 4 Horizon 2020 PuMMA: Studies Considering Reprocessing of 40-45 %Pu Fast Reactor MOx ........ 41 Chris Maher1*, Nathalie Chauvin2, Francisco Álvarez3, Martin Giraud 2, Jessica Gunning1, Victoria Hayter1, Rebecca Sanderson1, Nathalie Reynier-Tronche2, Eva de Visser Týnová4. ................................................................................................................................... 41 First-Principles Study of a New TODGA Degradation Compound ................................................................ 43 Lucas Zubillaga-Maharg1,2,3, Iván Sánchez-García4, Hitos Galán4, J. Manuel Perlado1,2, Emma del Rio1,2 ..................................... 43 Demonstration of the Single Cycle Am(III) Separation AmSEL Process in Laboratory-scale Annular Centrifugal Contactors .......................................................................................................................................... 45 Andreas Wilden1,*, Fynn S. Sauerwein1, Vincent Vanel2, Andreas Geist3, Giuseppe Modolo1 ................................................................... 45 Flowsheets for the Validation of the Reference AmSEL System ..................................................................... 47 Vincent Vanel1,*, Andreas Wilden2, Giuseppe Modolo2, Andreas Geist3, Marc Montuir1 .............................................................................. 47 Novel Water-soluble and CHON-compliant Ligands for Selective Americium Separation from PUREX Raffinate ............................................................................................................................................................................... 49 Elena Macerata,1 Alberto Arici,1 Giulia Firenze,1 Letizia Di Matteo,1 Fabrizio Piromalli,2 Andrea Salomone,1 Gabriele Magugliani,1 Eros Mossini,1 Mario Mariani,1 Alessandro Sacchetti2 .............................................................................................................................. 49 Extraction and Speciation Studies of New Diglycolamides with Varying Alkyl Chains for Selective Americium Partitioning ........................................................................................................................................ 51 Filip Kolesar [1,2], Karen Van Hecke [2], Ken Verguts [2], Cécile Marie [3], Laurence Berthon [3], Koen Binnemans [1], Thomas Cardinaels [1,2] ............................................................................................................................................................................................................................. 51 Selective Americium Separation: New insights into the Complexation of SO3-Ph-BTBP with Trivalent f-Elements .................................................................................................................................................................... 53 Fynn S. Sauerwein1,*, Thomas Sittel2, Andreas Geist2, Andreas Wilden1, Giuseppe Modolo1.................................................................... 53 Purification of Neptunium and Plutonium by Ion Exchange for Plutonium-238 Production at Oak Ridge National Laboratory ............................................................................................................................................ 55 Lætitia H. Delmau, David W. DePaoli, Brad N. Tinker, Dennis E. Benker, and Robert M. Wham ............................................................ 55 Experimental Studies and Molecular Modeling of the Physico-chemical Properties of Pure Monoamides Extractants ......................................................................................................................................................... 57 Abderrazak Masmoudi, Dominique Guillaumont, Philippe Guilbaud .................................................................................................................... 57 Development of Integrated Actinide Chemistry Application, AACE, for Acceleration of Actinide Chemistry Experiments ............................................................................................................................................................. 59 Masahiko Nakase1,2, Takahiro Nishihara1,2, Fauzia Hanum Ikhwan1, Tomohiro Okamura1,2, Kota Matsui3 ...................................... 59 Actinide Oxide Dissolution in Tributyl Phosphate ...................................................................................................... 61 Jarrod M. Gogolski, Chelsea M. Goetzman, Robert J. Lascola, Tracy S. Rudisill ............................................................................................... 61 Direct Extraction of Uranium from Used Nuclear Fuel with DEHiBA .............................................................. 63 Gabriel B. Hall, Nathan P. Bessen, Daria Boglaienko, Gregg J. Lumetta ............................................................................................................... 63 New Monoamide Based Extractants for U(VI) and Pu(IV) Efficient Separation ................................... 65 Cécile Marie,1 Pape Diabate,1 Audrey Beillard1 .......................................................................................................................................................................... 65 Efficient Manufacture of DEHiBA through Industry 4.0 .......................................................................................... 67 Tom Shaw, Prof. Bruce Hanson, Prof. Richard Bourne ......................................................................................................................................................... 67 Interinstitutional Study of the New EURO-GANEX Process Resistance by Gamma Irradiation Test Loops .......................................................................................................................................................................................... 69 5 Ivan Sanchez-Garcia,1 Xavier Heres,2 Dean. R. Peterman,3 Hitos Galan,1 Sylvain Costenoble,2 Sylvain Broussard,2 Johann Sinot,2 Travis S. Grimes,3 Kash Reid Anderson,3 Santa Jansone-Popova,4 Maria Chiara Gullo,5 Alessandro Casnati,5 Andreas Wilden,6 and Andreas Geist7 ...................................................................................................................................................................... 69 Americium Separation Processes Developed within the European PATRICIA Project ..................... 71 Christian Sorel,1 Andreas Geist,2 Giuseppe Modolo,3 Laure Ramond,1 Christian Ekberg,4 Hitos Galán,5 Bruce Hanson,6 Gregory Leinders,7 Elena Macerata,8 Cécile Marie,1 Petra J. Panak,2,9 Mark Sarsfield,10 Dan Whittaker,10 Andreas Wilden3 ........................................................................................................................................................................................................................................................................................71 Radiolytic Stability of Metal (IV) Phosphonate Sorbents Designed for Minor Actinide- Lanthanide Separations ........................................................................................................................................................... 73 Taren Cataldo1, Jessica Veliscek-Carolan2, Nicholas Bedford1, Sophie Le Caër3 ........................................................................................... 73 Optimization of Minor Actinides Recovery Conditions by Combining Mathematical Analysis and Process Simulation ............................................................................................................................................................ 75 Yuichi Sano1, Akane Kojima2, Tomoyuki Yajima2, Yoshiaki Kawajiri2 ......................................................................................................................... 75 Recent Results from Lab Scale Testing of Advanced Aqueous Separation Processes for the Future Recycling of Spent Nuclear Fuels ........................................................................................................................ 77 Robin Taylor1, Dan Whittaker1, Mark Sarsfield1, Michael Carrott1, Billy Keywood1, Rebecca Sanderson1, David Woodhead1, Kate Wallace1, Hongyan Chen2, Clint Sharrad3 ......................................................................................................................................................................... 77 Research on Sustainable Nuclear Energy Use with Actinide Management: Scenario Study on High-Level Waste Generation with MA Separation and Intermediate Storage Technology Implementation ............................................................................................................................................................................. 79 Tomohiro Okamura1, Ryo Takahashi2, Takashi Shimada2, Nakase Masahiko1, Tomoo Yamamura3 and Kenji Takeshita1 ...................................................................................................................................................................................................................................................................................... 79 Recovery of Strategic High-Value Fission Products from Spent Nuclear Fuel during Reprocessing .................................................................................................................................................................................... 81 Alistair F. Holdsworth[1], Harry Eccles[2], Kathryn George[1], and Clint A. Sharrad[1] ............................................................................... 81 Demonstration of Advanced Voloxidation and Direct Extraction Using Irradiated UO2 ................. 83 Peter Tkac1, Michael Kalensky1, Sergey Chemerisov2 ........................................................................................................................................................... 83 Zirconium Molybdate Rinsing With Carbonate: From R&D to Industrialization in the La Hague Plants ..................................................................................................................................................................................................... 85 Nicolas Golles1*, Céline Quenault2, Pierre Sarrat3, Sandrine Jakab-Costenoble3, Bénédicte Arab-Chapelet3, Nicolas Vigier4, Jean-Luc Emin4 ............................................................................................................................................................................................................................... 85 ACTINIDE MATERIALS AND NUCLEAR FUELS .............. 88 The Potentials of Nano-Scaled Actinide Dioxides ................................................................................................... 89 Olaf Walter, Karin Popa ............................................................................................................................................................................................................................. 89 Influence of Uranium Oxide Nature on MOX Fuel Fabrication Process ....................................................... 91 F. Sauvaire1, J-B. Parise1, A. Delevaque1, A. Selmi1, R.Delville2, T. Genevès1, V.Garat1 ......................................................................................... 91 New Insights in the Structural-Redox Chemistry of Cr, Mn, Fe and V doped-UO2 Nuclear Fuel Materials ............................................................................................................................................................................................. 93 G. L. Murphy [1], Philip Kegler [1]. Robert Gericke [2], S. Gilson [2], Martina Klinkgenberg [1], Daniil Sirochii [1], Andrey Bukaemskiy [1], Kristina Kvashnina [2], Voloydmyr Svitlyk [2], Christop Henig [2], Peter Kaden [2] and Nina Huittinen [2]................................................................................................................................................................................................................................................................................ 93 6 ESEM-Monitored Dissolution of (U,Th)O2 Heterogeneous Mixed Oxides for Spent Fuel Modeling ................................................................................................................................................................................................................... 95 L. Claparede [2], C. Hours [1], I. Viallard [3], N. Reynier-Tronche [1], N. Dacheux [2]................................................................................... 95 Defect Chemistry, Thermal Oxidation, and Thermodynamics of metal-doped UO2 ....................... 97 Xiaofeng Guo1, Juejing Liu1, Shinhyo Bang1, Lorenzo Callejon1,2, Romain Sicard1,2, Sam Karcher1, John McCloy1, Hongwu Xu3, Arjen van Veelen3, Nicolas Clavier2, Nicolas Dacheux2 ............................................................................................................................................. 97 Heat Capacity Measurements of Self-Damaged Mixed Actinide Oxides: a Method to Assess Defects in Spent Fuels ................................................................................................................................................................ 99 Thierry Wiss1, Rudy Konings1, Dragos Staicu1, Alessandro Benedetti1, Jean-Yves Colle1, Emilio Maugeri2, Zeynep Talip2, Emanuele De Bona3, Oliver Dieste4, Gianguido Baldinozzi5, Christine Guéneau6 ............................................................................................ 99 Conversion of U(VI) and Pu(IV) by Peroxide Precipitation and Hydrothermal Treatment ........... 101 L. Mullera,b, P. Estevenona, C. Tamaina, N. Dacheuxb, N. Clavierb.................................................................................................................................... 101 Densification Study of Cr-doped UO2 Fuel Pellets with Addition of Fission Products Surrogates ..................................................................................................................................................................................................................103 Antonin De Azevedo1, Fabienne Audubert1, Nathalie Moncoffre2, Jacques Lechelle1 ................................................................................ 103 Hydrothermal Reducing Conversion of Uranium(VI) Oxalate into Oxides ............................................ 105 S. Benarib 1, M. Munoz 1, I. Kieffer 2, J. L. Hazemann 3, N. Dacheux 1, N. Clavier 1 ................................................................................................ 105 Actinide Thioamidates as Precursors For Actinide Sulfide Nanomaterials ........................................... 107 Sheridon N. Kellyab, Dominic R. Russoab, Appie A. Petersonb, Erik T. Ouellettea, Michael A. Boreena, Jacob A. Bransonab, S. Olivia Guntherb, Patrick W. Smithb, John Arnolda, Stefan G. Minasianb. ..................................................................................................................107 Synthesis of PuO2 and (U,Pu)O2 Solid Solution by Citric Acid Assisted Combustion Synthesis109 A. Hautecouverture [1,2], P. Estevenon [1], C. Rey [2], X. Deschanels [2] ............................................................................................................ 109 Phase Separation in Fluorite-Related U 1– y Ce y O 2– x : New Insights via Variable Temperature Neutron Diffraction ........................................................................................................................................................................ 111 David Simeone1, Xavier Deschanels2, Philippe Garcia3, Maxim Avdeev4, Timothy Ablott5, Gordon J Thorogood6 .................. 111 Fabrication and Dissolution of Americium Plutonium Oxide Fuels .............................................................. 113 Eva de Visser-Týnová, Jessica Bruin, Frank Oud................................................................................................................................................................... 113 Fission Products Speciation in Irradiated MOx Fuel During Interim Storage Accidental Scenarios ............................................................................................................................................................................................ 115 R. Caprani1, Ph. Martin1, J. Martinez1, D. Prieur2, F. Lebreton1, K. Kvashnina2, E. Bazarkina2, D. Menut3, N. Clavier4 ....................... 115 Synthesis and Characterisation of CeO2 and PuO2 Pellets with Representative Microstructure for General Purpose Heat Sources .................................................................................................................................... 117 Jérémie Manaud, Walter Bonani, Jacobus Boshoven, Marco Cologna, Michael Holzhäuser, Karin Popa, Sorin-Octavian Vălu, Daniel Freis ............................................................................................................................................................................................................................................. 117 Incorporation of Fission Products into Oxide Nuclear Fuel: Towards a New Paradigm? ............... 119 L. Desgranges ................................................................................................................................................................................................................................................... 119 Quantification of the Morphology and Roughness of Oxide Powder Particles In Relation to their Manufacturing History and Flow Properties .................................................................................................... 121 Christophe D’Angelo1, Solène Bertolotto1, Anne-Charlotte Robisson1 and Christelle Duguay1 ............................................................ 121 Field Assisted Sintering of UO2 Based Nuclear Fuels ............................................................................................. 123 R.W. Harrison1, J. Morgan1, J. Buckley1, T. Abram1, D. Pearmain2, S. Bostanchi2, C. Green2, R. White2, D. Goddard3, N. Barron4 ....................................................................................................................................................................................................................................................................123 7 Characterization of the Phases Formed During the High Temperature Oxidation of (U,Pu)O2 Mixed Oxides.................................................................................................................................................................................... 125 Priscilla Berenguer-Besnarda,b, Philippe Martina , Loïc Marchettia, Loïc Favergeonb .................................................................................. 125 Fundamental Insights into Defect Generation and Transport Phenomena at Grain Boundaries in Nuclear Fuel ................................................................................................................................................................................ 127 S. Finkeldei1, J. Proctor1, O. Lori1, S. Dillon1, J. White2, Y. Wang2, M. Cooper2, D. Andersson2....................................................................... 127 Impact of Ru, Rh, Pd and Mo Metallic Particles on the Dissolution Kinetics of UO2. .......................... 129 Kaczmarek T.1, Szenknect S.1, Le Goff X.1, Podor R.1, Dacheux N.1 .................................................................................................................................. 129 Thermal Oxidation and High Temperature Structures of Uranium Carbide: in situ X-Ray Diffraction Studies ........................................................................................................................................................................ 131 Emma C. Kindall, Natalie S. Yaw, Juejing Liu, Malin C. Wilkins, Sam Karcher, Hongwu Xu, Arjen van Veelen, Josh T. White, John McCloy, Xiaofeng Guo ................................................................................................................................................................................................... 131 WASTE CONDITIONING AND GEOLOGICAL REPOSITORY ................................................................. 134 REGAIN project – Recycling of Zirconium from Nuclear Hulls .........................................................................135 Mathilde Guilpain, Isabelle Hablot, Bertrand Morel ........................................................................................................................................................... 135 Successful Lasergrammetry Operation in an ATALANTE Hot Cell: First Step for Deploying Digital Technologies on Hot Cells In Operation ........................................................................................................................ 137 Caroline Chabal (1)*, Michaël Gauthier (1), Julien Delrieu (1), Thibaud Durand (1), Christian Père (2)** ................................................... 137 Effects of Radiolysis Products and Acidic Media on the Aggregation Behaviour of Nuclear Fuel Debris Nanoparticle Simulants via Stochastic Simulations ............................................................................139 Miguel Pineda1, Cong Chao1, Yiwei Zhang1,2, Takehiko Tsukahara2, Panagiota Angeli1, Eric S. Fraga1 ............................................ 139 Insights into Glass Alteration Mechanisms from the Study of Long Term Burial Experiments .. 141 Clare L. Thorpe, Garry Manifold, Rachel Crawford and Russell J. Hand ............................................................................................................... 141 Impact of Lanthanide and PGM Elements on the Chemical Durability and Surface Modifications during the Leaching Tests of FP Doped Pellets Mimicking Interim Repository....143 P.H. Imbert, L. Claparede, S. Szenknect, N. Dacheux ........................................................................................................................................................... 143 The Impact of Hot Isostatic Pressing on U Speciation and Local Coordination in Simulant Pu Ceramic Wasteforms ............................................................................................................................................................... 145 Aidan Friskney1, Ewan R. Maddrell2, Claire L. Corkhill3 and Lewis R. Blackburn1 .............................................................................................. 145 Impact of Gamma Dose Rate on the Alteration of Nuclear Glass in Geological Disposal Conditions ......................................................................................................................................................................................... 147 M. Taron [1, 2], H. Aréna [1], F. Chupin [1], K. Ressayre [1], R. Podor [2], M. Tribet [1], S. Peuget [1] .....................................................147 Low-Temperature Condensation and Solidification of Radioactive Liquid Waste by Freeze Drying .................................................................................................................................................................................................. 149 Akihiko Kajinami1, Sou Watanabe2 ................................................................................................................................................................................................... 149 Investigation of Cement-Based Materials with Dihydrogen Sequestration Properties ................. 151 H. Danis1), C. Cau Dit Coumes1), P. Antonucci1), I. Pointeau2), T. Cordara2), N. Macé 3), S. Savoye3) ............................................... 151 Microwave Plasma-Assisted Combustion of Waste Organic Solvents ...................................................153 8 Shimpei Ohno1, Atsushi Sakamoto1, Sou Watanabe1, Masahiro Nakamura1, Tsuyoshi Yamamoto2 and Ryou Tanaka3 . 153 Search for a Cement Matrix for ITER Beryllium Radwaste Conditioning ................................................. 155 Laflotte R.1), Cau Dit Coumes C.1), Cannes C.2), Delpech S.2), Haas J.1), Rivenet M.3) ....................................................................................... 155 Repercussions of Solubility for the Conditioning of Fission Products and Minor Actinides in Borosilicate Glasses ................................................................................................................................................................... 157 L. Campayo, I. Giboire, S. Schuller, E. Régnier, D. Perret .....................................................................................................................................................157 Progress towards the Immobilisation of the UK Plutonium Inventory in Titanate Ceramics .... 159 Lewis R. Blackburn1, Amber R. Mason1, Laura J. Gardner1, Claire L. Corkhill2 ..................................................................................................... 159 Elaboration and Characterization of Iodate and/or Carbonate-Doped Apatites for Long-Lived Radionuclides Conditioning .................................................................................................................................................. 161 Olivier Dautaina,b, Céline Cau dit Coumesb, Christophe Droueta, David Grossina,Lionel Campayob, Christèle Combesa ...................................................................................................................................................................................................................................................................................... 161 The Effect of Cation Substation and Valency on Formation Energetics of Brannerite Ceramics for Nuclear Waste Applications ......................................................................................................................................... 163 Natalie S. Yaw, Chris Dixon Wilkins, Nicolas Dacheux, John McCloy, Xiaofeng Guo .................................................................................. 163 Densification of Mesoporous Silicas Induced by Radiation Damage - New Perspectives for the Treatment of Radioactive Effluents ................................................................................................................................. 165 Jun Lin*, Clara Grygiel**, Sandrine Dourdain*, Yannick Guari***, Cyrielle Rey*, Jérémy Causse*, Olaf Walter****, Xavier Deschanels* ..................................................................................................................................................................................................................................................... 165 A Historical Overview of Corroded Microstructures and Present-day Best Practices. .................. 167 Mir Anamul Haq ............................................................................................................................................................................................................................................ 167 Compared Radiation Stability of Mesoporous Silica and Nuclear Glass Alteration Gels ........... 169 Pierre De Laharpe*1, Xavier Deschanels1, Sylvain Peuget2, Helene Arena2, Melanie Taron1,2, Jun Lin1, Bertrand Siboulet1, Jean-Marc Delaye2 ...................................................................................................................................................................................................................................... 169 Insights into the Structural and Redox Chemistry of Cr-doped (Ln,U)O2 Materials ......................... 171 Daniil Shirokiy,1 Maximilian Henkes,1 Andrey Bukaemskiy,1 Kristina O. Kvashnina,2 Martina Klinkenberg,1 Philip Kegler,1 Dirk Bosbach,1 and Gabriel L. Murphy,1 ............................................................................................................................................................................................. 171 Simulating Auto-Irradiation of Glass Using External Irradiation Beams: Impact on Glasses Structure and Properties ......................................................................................................................................................... 173 M. Taron 1, 2, H. Aréna 1, C. Gillet 1, F. Perrudin 1, R. Podor 2, M. Tribet 1, S. Miro 1, S. Peuget 1 ........................................................................... 173 The Influence of pH, Ionic Strength and Temperature on Cs, Ba, Co, and Eu Sorption on Biotite - Experiments and Modelling ............................................................................................................................................... 175 Pawan Kumar, Stellan Holgersson, Christian Ekberg ........................................................................................................................................................175 Colloids Pose an Enhanced Transport Risk of Uranium in Saturated Porous Media: A Challenge for Immobilization Remediation of Uranium Contaminated Site ................................................................ 177 Duoqiang Pan, Xiaoyan Wei, Xinyi Shi, Weixiang Xiao, Zhen Xu, Wangsuo Wu .............................................................................................. 177 Processes Driven by Iron Reducing Bacteria on Technetium Immobilization ..................................... 179 Cardaio I., Müller K., Cherkouk A., Stumpf T., Mayordomo N. .........................................................................................................................................179 SAFEGUARDS AND ANALYTICAL CHEMISTRY ........... 182 Real-Time and Automated Process Control via On-Line Monitoring ...................................................... 183 9 Amanda M. Lines,1* Poki Tse,1 Nathan Bessen,1 Thomas Serrano,1 Alyssa Espley,1 Gilbert Nelson,2 Gabriel B. Hall,1 Jarrod R. Allred,1 Gregg J. Lumetta,1 and Samuel A, Bryan1 ................................................................................................................................................................... 183 Photonic Lab-On-A-Chip, a Versatile and Powerful Tool for R&D Studies on Spent Fuel Reprocessing ................................................................................................................................................................................. 185 Fabrice Lamadie1*, Elodie Mattio1, Manuel Miguirditchian1, Amanda M. Lines2, Samuel A. Bryan2, Hope E. Lackey2, Fabrice Onofri3, Isaac Rodriguez-Ruiz4 .......................................................................................................................................................................................... 185 Real-Time Solution Analysis in Microfluidic Devices using Optical Spectroscopy ........................... 187 Samuel A, Bryan,1* Hope E. Lackey,1 Gilbert L. Nelson,2 Job M. Bello,3 Fabrice Lamadie,4 and Amanda M. Lines1 ...................187 The Joint Research Centre’s Expertise in Nuclear Safeguards Sample Analysis ............................. 189 A.M. Sánchez Hernández, R. Buda, K. Casteleyn, L. Commin, F. D’Amati, J. Horta, A. Le Terrier, A. Muehleisen, S. Stohr, H. Schorlé, M. Toma, M. Vargas Zuñiga, D. Wojnowski, J. Zsigrai, K. Mayer ............................................................................................................... 189 A New Plutonium Metal Certified Reference Material at CETAMA: the MP4 Standard ..................... 191 S. Picart1, M. Crozet1, Y. Davrain1, C. Bertorello1, G. Canciani1, C. Rivier1, D. Cardona2, G. Legay2, G. Bailly2, N. Caussignac2,C. Zeleny2, A.Quemet1, S. Baghdadi1, C. Maillard1, V. Dalier1, L. Montreuil1, S. Jan1, S. Mialle3, C. Cruchet3, H. Isnard3, W. Pacquentin3, F. Doreau1, P. Estevenon1, J. Lorino1, L. Picard1, S. Richter4, Y. Aregbe4, A. M. Sanchez Hernandez5, H. Schorle5, R. Buda5, U. Repinc6, M. Kohl6, J. Hiess6, G. Duhamel6, M. Sumi6. ................................................................................................................................... 191 Development of Uranium Oxide-based Reference Microparticles for Particle Analysis in Nuclear Safeguards ...................................................................................................................................................................193 Stefan Neumeier1, Shannon Potts1, Philip Kegler1, Giuseppe Modolo1, Martina Klinkenberg1, Simon Hammerich2, Dirk Bosbach1, Irmgard Niemeyer1 ............................................................................................................................................................................................................... 193 Laser Ablation- ICP-MS Method Development for a Self-Consistent Calibration in Post Irradiation Examination of Spent Fuels ......................................................................................................................... 195 Peter Zsabka1, Kyle Johnson2 ............................................................................................................................................................................................................... 195 Burnup Determination of Irradiated U-Mo Alloy Fuel by 148Nd Monitor Method .................................. 197 Hyejin Cho, Namuk Kim, Yang-Soon Park, Minjae Ha, Tae-Hong Park, Hye-Ryun Cho, Jai Il Park...................................................197 On L-edges X-ray Emission Spectroscopy as a Tool to Study Actinide’s Electronic Structure: The Case of Uranium in UxOy Compounds ................................................................................................................. 199 P. Silvenoinnen, I. Prozheev, R. Bes.................................................................................................................................................................................................... 199 PYROCHEMISTRY AND CHEMISTRY FOR MOLTEN SALTS ........................................................................... 202 Overview of Plutonium Pyroprocessing By-Products Management ....................................................... 203 G. Bourgès, S. Faure, O. Lemoine, D. Cardona - Barrau .................................................................................................................................................. 203 Spent Fuel Reprocessing for Molten Salts Fast Neutron Reactors ............................................................. 205 A. Handschuh1, P. Ryckewaert1, P. Baron1, S. Delpech2, C. Cannes2, D. Lambertin1, T. Kooyman1 ........................................................ 205 Pyrochemical Treatment for Molten Salt Nuclear Reactor ..............................................................................207 Joelle Costantine,1 Davide Rodrigues,1 Céline Cannes,1 Elisa Capelli,2 Bertrand Morel,2 Sylvie Delpech1 ................................... 207 Feasibility of Lanthanide Extraction Assisted by Electrolysis on Li-Bi Liquid Cathode in Molten Fluorides. .......................................................................................................................................................................................... 209 Pierre Chamelot, Mathieu Gibilaro and Laurent Massot ............................................................................................................................................... 209 10 Molten Salts and Pyrochemical Processing Progress at the UK’s National Nuclear Laboratory ................................................................................................................................................................................................................... 211 Ruth Carvajal-Ortiz, Mike J Edmondson, Moya Hay ............................................................................................................................................................ 211 Synthesis of Actinide Chlorides as Fuel for Fast Molten Salt Reactor ........................................................ 213 P. Chevreux1, M. Duchateau1, G. Serve1, M. Pons1. ....................................................................................................................................................................213 Molten Salt Spectroelectrochemistry in Chloride Based Eutectic Systems with Uranium .......... 215 Jessica A. Jackson, Nicole Hege, Jacob Tellez, Jenifer Shafer .................................................................................................................................. 215 Influence of Nitrogen on Uranium Metal Stability in Molten LiCl-KCl ......................................................... 217 Théo Caretero1, Laurent Massot1, Mathieu Gibilaro1, Jérémie De Marco2, Mehdi Arab2, Pierre Chamelot1 .................................. 217 ACTINIDE AND FISSION PRODUCTS CHEMISTRY ..... 220 Elucidating the Radiation-Induced Redox Chemistry of Plutonium Under Used Nuclear Fuel Reprocessing Conditions ........................................................................................................................................................ 221 G. P. Holmbeck, Amy E. Kynman, Jacy K. Conrad, and Travis S. Grimes ............................................................................................................. 221 How Plutonium “Brown” Peroxo Complex emerges from Aerated Electrolysis Experiments .... 223 Quentin Hervy1, Richard Husar1, Thomas Dumas1, Philippe Guilbaud1, Matthieu Virot2, Philippe Moisy1 ...................................... 223 The Redox Chemistry of [MIV/III(3,4,3-LI(1,2-HOPO))]0/- Complexes in Acidic Aqueous Media 225 Jeffrey R. McLachlanab*, Andrae A. Tabbsa, Alex C. Rigbya, Joshua J. Woodsa, Rebecca J. Abergelab* ...................................... 225 Chronicles of Peroxide Plutonium Species: Structural Characterization of New Pu(IV) Green Peroxide ............................................................................................................................................................................................. 227 J. Margate [1], M. Virot [1], S. Bayle [1], D. Menut [3] T. Dumas [2], E. Dalodière [1], C. Tamain [2], P. Estevenon [2], P. Moisy [2], S. I. Nikitenko [1] ......................................................................................................................................................................................................................227 Development of Water-Compatible N3O2-Pentadentate Planar Ligands for Uranium Harvesting from Seawater ................................................................................................................................................... 229 Koichiro Takao,1 Takumi Mizumachi,1 Minami Sato,1 Koma, Ito,1 Ryuto Nabata,1 Masashi Kaneko,2 Tomoyuki Takeyama,1,3 Satoru Tsushima4,5 ....................................................................................................................................................................................................................................... 229 Complexation and Solvent Extraction Properties of the N, N, N’, N’-tetraethyl-1,10- phenanthroline-2,9-Diamide Extractant with Ln(III) and An(III) ................................................................... 231 Emma M. Archer†, Jocelyn M. Riley†, Felipe A. Pereiro╪, Elizabeth B. Flynn†, Jacob P. Brannon†, Stosh A. Kozimor╪, Jessica A. Jackson†, Jenifer C. Shafer†, and Shane S. Galley† ..........................................................................................................................................................231 Speciation of Uranium(VI) with Amido-Phosphonate Ligands in Organic Phase and at the Solid/Liquid Interface Studied by Molecular Dynamics ....................................................................................233 Diego Moreno Martinez, Dominique Guillaumont, and Philippe Guilbaud ..................................................................................................... 233 Probing the Metal Ion-Ligand Interaction in An(III) and Ln(III) Complexes: An Overview about Recent Advancements ........................................................................................................................................................... 235 Thomas Sittel,1 Patrik Weßling,1,2 Andreas Geist,1 Petra J. Panak1,2 ........................................................................................................................... 235 Reactivity of Actinides Mono-Cations with NH3 in Gas Phase: A Study Using ICP-MS and Quantum Chemistry ................................................................................................................................................................. 237 M. Goujet [1], A. Quémet [1] and D. Guillaumont [1] .......................................................................................................................................................... 237 11 Crystal Growth of New Uranium and Transuranic Phases via High Temperature Solution and Mild Hydrothermal Methods: Exploration of New Materials as Potential Nuclear Waste Forms ................................................................................................................................................................................................................ 239 H.-C. zur Loye,1,2,3,* T. K. Deason1,2,3, A. T. Hines,1,2 H. Tisdale,1,2 T. M. Besmann,1,2 J. Amoroso,1,3 D. P. DiPrete,1,3 ................................ 239 Molecular and Crystal Structures of Pu(IV)-Nitrato Complexes with Double-Headed 2- Pyrrolidone Derivatives in HNO3(aq)................................................................................................................................ 241 Ryoma Ono1, Tomoyuki Takeyama1, Robert Gericke2, Juliana März2, Tamara Duckworth2, Satoru Tsushima2,3, Koichiro Takao1 ..................................................................................................................................................................................................................................................................... 241 The Adsorption of Pu(IV) in the Presence of Cesium Phosphomolybdate, Barium-Strontium Nitrate, Zirconium Molybdate and Zirconium Hydrogen Phosphate ....................................................... 243 Joshua Turnera, Christian Whitea, Jessica Blenkinsopa, Barbara Dunnetta, Adam Bragga, Dan Whittakera, Richard Lynnb ................................................................................................................................................................................................................................................................................... 243 Performance and Design of HotXAS: the Future In-House XAS Apparatus at Atalante .............. 245 S. Orlat1,2, R. Bes2, F. Tuomisto2, P. Martin1, P. Moisy1 .............................................................................................................................................................. 245 Investigation of the Microcosmic Dynamics Behaviors of Hydrogen and Oxygen in Plutonium Oxide via Ab Initio Molecular Dynamics Simulations ..........................................................................................247 Jun Tang ............................................................................................................................................................................................................................................................ 247 12 PLENARY TALKS 13 Global Overview on the Nuclear Fuel Cycle Backend and IAEA Related Activities Clément Hill, Amparo Gonzalez-Espartero International Atomic Energy Agency, Vienna International Centre, PO Box 100, Vienna, Austria, 1400 (c.hill@iaea.org) One of the main concerns of the population regarding nuclear power is the safe management of the Spent Nuclear Fuel (SNF) after discharged from the reactors. There are however technical solutions, implemented since the inception of nuclear power in the 50’s, and SNF has been safely managed for more than 60 years, worldwide. Currently, 417 nuclear power reactors operate in 31 countries. SNF is discharged at a rate of about 10,000 tHM/year. As of the end of 2022, the accumulated SNF inventory that has been globally discharged since the deployment of nuclear power amounts to approximately 430,000 tHM, of which around 70% are currently stored (under wet and dry conditions), waiting for further steps or final disposition, and the rest has been reprocessed (mostly for recycling valuable fissile materials). The lifetime extension of some nuclear reactors is contributing to an increase in the amount of stored SNF. Understanding the behaviour of SNF in various storage systems, as well as the ageing and degradation mechanisms of storage structures, systems and components, remains vital to ensure that SNF can continue to be stored safely and subsequently transported to disposal or reprocessing facilities, even after long term storage. As spent fuel disposal programmes are progressing and approaching the final stages of construction in some Member States, there has been an increase in the number of preparation activities, such as the development of characterization programmes. Continuation of such efforts is particularly important when considering that greater reactor efficiencies have been achieved through the production of SNF with higher initial enrichments and higher burnups, leading to increases in thermal outputs and potentially higher risks of cladding embrittlement that may impact subsequent SNF management steps. Despite an overall reduction in global spent fuel reprocessing capacity (after the closure of reprocessing plants in the United Kingdom), there is an increasing interest in the development of advanced recycling technologies. As new fuel designs are envisaged for both existing fleets of nuclear reactors and those of future advanced reactor designs (including SMRs), which may lead to potentially different behaviours in SNF management, innovative SNF management solutions will need to be sought to allow for their timely deployment. International collaboration and partnership will therefore be paramount for their success. The main goal of nuclear scientists, engineers, operators, and regulators, today, is thus to continue improving, developing and licensing new technologies to address current and future challenges of the nuclear fuel cycle backend, so as to ensure the nuclear fuel cycle sustainability, through advanced nuclear energy systems that can preserve natural resources and reduce the burden of the nuclear waste generated. Integration of new and innovative fuel cycles with existing fuel cycles is an important undertaking to address current energy supply challenges and ensure the sustainable, safe and secure development of nuclear power. While already implemented in some countries, initiatives to address advanced reactors’ spent fuel and radioactive waste management, in an integrated manner, are starting to be discussed and developed in several countries. The International Atomic Energy Agency (IAEA) supports sustainable, safe, secure, reliable and economic nuclear fuel cycles associated with current and future generations of nuclear power reactors, by providing its 178 Member States with relevant technical information (guidance) based on operational experience and best industrial practices, through: - The coordination of international research activities, via Coordinated Research Projects [1-3] (CRPs), the final reports of which compile research findings, and policy and strategy approaches in Member States; - The organization of technical events, workshops and international conferences to provide Member States with fora for sharing technical information, such as: o The Technical Meeting on “Challenges, Gaps and Opportunities for the Management of SNF from SMR Technologies”, that gathered more than 100 participants from 32 countries in September 2022, the Proceedings of which were published last December [4], o The International Conference on Spent Fuel Management from Nuclear Power Reactors, to be held in Vienna, on 10-14 June 2024; mailto:c.hill@iaea.org 14 - The publication of technical reports [5,6] or guides (including interactive publications [7]), compiling relevant information and best practices; - The management of specific databases; and - The development of e-learning materials [8] and infographics (Fig.1) for capacity building in Member States. Fig.1. Inventory of SNF around the world (amount of SNF discharged, reprocessed and stored by country and by storage system) and Dry Storage technologies. The presentation will provide an overview on the issues of the backend of the nuclear fuel cycle, worldwide, and on the IAEA programmatic activities in support of its Member States under the sub-Programme on the Management of Spent Fuel from Nuclear Power Reactor and Radioactive Material Transportation. References: [1] CRP T13020 on “Spent Fuel Research and Assessment (SFERA)”: https://www.iaea.org/projects/crp/t13020. [2] CRP T13019 on “Performance Assessment of Storage Systems for Extended Durations (PASSED)”: https://www.iaea.org/projects/crp/t13019. [3] CRP T13021 on “Challenges, Gaps and Opportunities for Managing Spent Fuel from Small Modular Reactors”: https://www.iaea.org/projects/crp/t13021. [4] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA-TECDOC-2040, Considerations for the Back End of the Fuel Cycle of Small Modular Reactors. Proceedings of a Technical Meeting, IAEA, Vienna (2023). [5] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA-TECDOC-1967, Status and Trends in Pyroprocessing of Spent Nuclear Fuels, IAEA, Vienna (2021). [6] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Nuclear Energy Series No. NF-T-4.11, Technical Approaches for the Management of Separated Civilian Plutonium, Vienna (2023). [7] INTERNATIONAL ATOMIC ENERGY AGENCY, Guidebook on Spent Fuel Storage Options and Systems: Guidebook on Spent Fuel Storage Options and Systems (iaea.org). [8] Course on Spent Fuel Management: Course on Spent Fuel Management | IAEA. https://www.iaea.org/projects/crp/t13020 https://www.iaea.org/projects/crp/t13019 https://www.iaea.org/projects/crp/t13021 https://nucleus.iaea.org/sites/connect/SFMpublic/SF%20Guide%20book/www/index.html#/reader https://nucleus.iaea.org/sites/connect/SFMpublic/SF%20Guide%20book/www/index.html#/reader https://www.iaea.org/online-learning/courses/561/course-on-spent-fuel-management 15 Status of the French Nuclear Fuel Cycle Program François Sudreau Energy Division, CEA, Centre de Paris-Saclay, 91191 Gif-sur-Yvette, France Context This presentation gives an update concerning the French nuclear fuel cycle R&D program. Currently, the French nuclear fuel cycle is based on plutonium monorecycling, that is to say that plutonium is recovered from UOx spent fuel and reused in MOX fuel. After irradiation, these MOX fuels are stored. They are not considered in France as nuclear waste as they contain valuable fissile materials, including around 6 % of plutonium, and as the feasibility of their reprocessing has been demonstrated since 70 tonnes of spent MOX fuels have already been reprocessed in France, at Marcoule and La Hague facilities. This strategy allows to save 10 % of natural uranium compared to an open cycle, avoids to have plutonium in the final waste to be stored in deep geological repository, decreases the net production of plutonium of the nuclear reactor fleet and allows to stabilize the UOx spent fuel inventory. On February 2022, President Macron expressed his wish that, in addition to pursuing the massive development of renewable energy sources, the lifepan of current reactors be extended, subject to the agreement of the ASN, whenever possible. In addition, a program for six EPR2 type reactors is launched as a first step to guarantee France's energy independence and achieve carbon neutrality by 2050. Moreover, French Government has launched a major Investment Plan for the Future called "France 2030". One objective of this plan is to promote the emergence, in France, of innovative nuclear reactors. In that plan, approximately €500 million are allocated to innovative reactor projects, based on a call for projects that has been published in March 2022. Research and innovation around breakthrough nuclear reactor concepts should thus provide new answers to the challenges specific to the nuclear industry, for example in terms of competitiveness, safety, security, closing the nuclear fuel cycle or reducing the volume and activity of radioactive waste. Thus, 11 projects (9 for fission reactors and 2 for fusion ones) were selected to be awarded, leading to a cumulative state support of 130 million euros. All these decisions have an important impact on the future French nuclear fleet and consequently on possible nuclear fuel cycles. In this field, in February 2024, the Nuclear Policy Council, chaired by President Macron, confirmed the main thrusts of French policy on the fuel cycle back end, combining reprocessing and reuse of spent fuel, with an objective of a closed fuel cycle. Following this decision, Bruno Le Maire, French Minister of the Economy, Finance and Industrial Sovereignty, announces in March 2024 (i) a program to extend the lifetime of the La Hague and Melox plants beyond 2040, (ii) the launch of studies for a new MOX fuel fabrication plant and (iii) the launch of studies for a new spent fuel processing plant by 2045/2050. Plutonium Multirecycling in LWR For the medium term, CEA, Orano, EDF and Framatome have launched a major R&D program and associated industrial feasibility studies for plutonium multirecycling in Light Water Reactor. Three strategic objectives have been defined. First, characterize the interest of multirecycling compared to monorecycling in terms of materials and waste management. Second, asses the industrial feasibility of the multirecycling in Light Water Reactors considering, on one hand, the impacts on the reactors in terms of safety and performances and, on the other hand, on the current and future backend facilities and logistics (particularly transportations). Third, evaluate the technical and economic impact of multirecycling compare to monorecycling. After several assessments and optimizations of fuels made with multi-recycled plutonium and core management strategies, a standard MOX fuel assembly, using a low fissile plutonium quality, called MOX-MR for “MOX-Multi-Recycling” has been defined. Industrial deployment scenarios have been simulated, considering a future French EPR fleet with a combined installed capacity of 40 or 50 GWe, and plutonium multirecycling from 2046. These simulations show that, until the end of the century (i) plutonium and uranium multirecycling in LWR could allow saving 40 % of naturel uranium compared to an open cycle, (ii) spent MOX fuel inventory could be stabilized and total spent fuel inventory reduced and (iii) plutonium inventory can thus be stabilized. 16 Closing the nuclear fuel cycle Beyond plutonium multirecycling in PWR, France maintains the objective of the full closure of the nuclear fuel cycle using Fast Reactors. The goals of this next step are (i) the full independence from natural uranium input using the depleted uranium stockpiles resulting from enrichment activities and (ii) the long-term stabilization of zero net production of plutonium and spent fuel. The reference path is to pursue the development of high-power Sodium-cooled Fast Reactor technology and deploy them, if necessary, by the end of the century. In the same time, in the frame of France 2030 plan, other concepts, particularly Advanced Modular Fast Reactors, such as SRF (sodium cooled), LFR (lead cooled) and MSR (molten salt) reactors, are considered by new comers to be part of the solution for the reduction of the MOX fuel inventory and even of the waste that today are destined to be stored in deep geological repositories. In this context, studies have been launched to explore the potential of fast neutron molten salt reactors for actinide transmutation. These reactors could be fed only with plutonium (in particular plutonium with a low isotopic vector) or with plutonium and minor actinides. An R&D on the priority items, with lower technological maturity, is thus in progress in the frame of the Fourth Investment Program for the Future started in 2022. Impacts on treatment and fabrication processes Today, with plutonium monorecycling, there is no industrial MOX fuel treatment and MOX fuel is manufactured with a 9 % plutonium content. With plutonium multirecycling in PWR, it would be necessary to treat MOX and MOX2 fuels and MOX fuel production would be multiplied by around three. Moreover, plutonium multirecycling in Fast Reactors would need to treat and manufacture MOX fuels that could have a plutonium content up to, at least, 25 % and probably 30 %, or even more. Thus, some challenges have to be overcome to implement plutonium multirecycling. Concerning MOX reprocessing, two main steps are considered in French R&D programs: dissolution and separation. For the dissolution step, voloxidation, which allows to turn irradiated fuel pellets into powders, and thus to increase surface contact and dissolution kinetics, is investigated theoretically and experimentally on fresh and irradiated MOx fuels. For the separation step, a program is performed to develop and industrialize a single cycle process without oxydoreduction reactions. Indeed, using monoamide solvents seems to be an interesting alternative to TBP due to their good stability, high selectivity and high recovery factor for uranium and plutonium. This high selectivity allows considering a simpler flowsheet with a single cycle for extraction and purification of plutonium and uranium. Moreover, in this process, partitioning is made by changing the acidity of the solution, avoiding using oxydoreduction and stabilisation agents. This will lead to a process well adapted to high plutonium contents, more compact, less expansive and easier to control than the current separation process. Conclusion To conclude, the use of nuclear energy will increase and diversify in the world in the next decades. Thus, it is necessary to preserve the natural resources recycling reusable fissile materials. France is evaluating a step-by-step approach starting with plutonium multirecycling in PWR, before reaching the full nuclear fuel cycle closure using Fast Reactors. In that way, France is performing a large R&D program, particularly on innovative reprocessing and fuel fabrication processes. At the same time, through collaborations between CEA, Orano, EDF, Framatome and CNRS, as well as with new emerging players, the value of advanced technologies is being assessed with a view to improving the management of plutonium and minor actinides. 17 Nuclear Fuel Recycle Activities in the Office of Nuclear Energy Stephen Kung Office of Nuclear Fuel Cycle, U.S. Department of Energy 19901 Germantown Road, Germantown, MD 20874-1207 This presentation will cover the nuclear fuel recycle research and development (R&D) activities supported by the U. S. Department of Energy, Office of Nuclear Energy (DOE-NE). DOE-NE conducts applied R&D on advanced fuel recycle technologies that have the potential to improve resource utilization and energy generation, reduce waste generation, and limit proliferation risk. DOE-NE focuses on developing advanced fuel recycling technologies and addressing fundamental materials separations and recovery challenges that present significant degrees of technical risks and financial uncertainties. DOE-NE stewards the capabilities and knowledge relied upon by policy makers to make informed decisions regarding nuclear fuel cycle options. Such decisions in turn rely on the development of efficient and economical separation methods that can accept the used nuclear fuel containing actinides and fission products to recycle selected actinides, recover valuable by-products, and deliver waste streams that are suitable for disposal. DOE-NE supports the development and demonstration of various recycling technologies to make available small quantities of high-assay low enriched uranium materials for advanced reactor fuel-fabrication R&D needs. 18 19 Future Fuel Cycles – a UK Perspective Paul Nevitt National Nuclear Laboratory, Chadwick House, Birchwood Park, Warrington, WA3 6AE, UK UK Civil Nuclear Roadmap and Current Perspective UK Government has set out a vision and plan for fuel cycle in the UK. The UK Civil Nuclear Roadmapa published earlier in 2024 by UK Government set out how the fuel cycle underpins both government’s net zero and national security objectives. The roadmap set out key aspects of future fuel cycles in the UK, As the UK embarks on a nuclear renaissance, the ability to deliver this in a way that continues to enable national security outcomes depends on the UK regenerating its domestic fuel cycle capabilities. This focuses on three key areas: front end fuel production, research and innovation in advanced uranium-based fuels, and maintenance and development of a skilled workforce. In addition, while the current inventory of spent nuclear fuel and radioactive waste is either stored on nuclear power station sites or at Nuclear Decommissioning Authority (NDA) sites such as Sellafield, the UK long-term strategy for finally disposing of the most highly active radioactive waste inventory is to develop an engineered Geological Disposal Facility (GDF). A process is well underway to identify a suitable site in which to develop a GDF that has suitable geology and the support of a local community. However, the first spent fuel is not expected to be placed into a GDF until the 2050s. Until then, there is sufficient interim storage for the inventory from UK legacy, existing nuclear fleet and currently planned future plants. Finally, the UK reprocessed spent fuel on an industrial scale from the 1950s to 2022. Commercial industrial scale reprocessing came to an end in the UK with the closure of the Thermal Oxide Reprocessing Plant (THORP) in 2018. There is currently no industrial scale reprocessing in the UK and no plans to restart reprocessing. Managing the UK Legacy It is clear that the journey from nuclear fuel cycle to waste management and environmental restoration is evolving in the UK. The NDA has responsibility for managing inventories of legacy nuclear materials including spent fuels and plutonium on behalf of the nationb. The NDA mission will span decades and takes billions in investment, it will require new plants, new processes, new stores, new disposal facilities. It is replete with technical challenges and opportunities. All needing highly multi-disciplinary teams with diverse skill sets. The strategic approach in the UK is focused around: consolidate, storage, disposition (right solution), safeguards assurance, counter-proliferation. In relation to spent fuels disposition, in 2012 the NDA and UK government took the decision to complete the reprocessing contracts and commit the remaining (predominantly) Advanced Gas Reactor (AGR) fuel, and future arisings, to long-term pond storage. This is considered the most cost-effective and viable strategy but means that the UK had to underpin this approach including the storage and disposal of around 5,000 tonnes of AGR fuel. Research and development to underpin and support this includes: corrosion trials of fuels, condition monitoring of fuels, disposability studies. The UK has also set out its approach to plutonium management through the NDA Strategyb. Plutonium has been produced at Sellafield since the early 1950s from the reprocessing of spent fuel from UK power stations and overseas utilities. There is around 141 tonnes owned by NDA and its customers. Most of it is stored as a powdered oxide form in storage cans but some of the material is in the form of residues and various powders, pellets and assemblies. Government’s objective is to put the plutonium beyond reach. This could be by reuse as Mixed Oxide Fuel (MOX) in nuclear reactors or as an immobilised product. This would put the material in a form which reduces the long-term security risks and burden during storage and is aligned with its ultimate disposal in a GDF. In addition, the plutonium science and support to can surveillance programme is critically important to reducing the risks with the long-term storage of plutonium. The NDA are continuing to progress the development of the reuse as MOX and immobilisation options. a https://www.gov.uk/government/publications/civil-nuclear-roadmap-to-2050 b https://assets.publishing.service.gov.uk/media/605cb82fd3bf7f2f112f0f84/NDA_Strategy_2021_A.pdf https://www.gov.uk/government/publications/civil-nuclear-roadmap-to-2050 https://assets.publishing.service.gov.uk/media/605cb82fd3bf7f2f112f0f84/NDA_Strategy_2021_A.pdf 20 Key strategic decisions have been made by NDA and UK government over the last decade to bring about the benefits of consolidation of fuels and materials and the completion of reprocessing. There will be considerable demand for R&D to support future strategy development and underpinning. For spent fuels much of this technical work is around approaches to storage, transport and ultimately disposal. The plutonium programme will last decades, and the supporting R&D programme is a major UK endeavour. Looking to the Future Finally, looking over the longer term towards future sustainable fuel cycles, the UK invested significantly in the ‘Nuclear Innovation Programme’. This included, the Advanced Fuel Cycle Programme (AFCP)c, led by the National Nuclear Laboratory, to develop the next generation of nuclear fuels and fuel cycles. AFCP marked the biggest public investment in future nuclear fission fuel cycle R&D in a generation in the UK – AFCP looked at the role of advanced nuclear fuels and fuel cycles for a Net Zero future (Figure 1). With UK Government committed to achieve carbon neutrality by 2050, AFCP aimed to lay the foundation from which the UK could maximise its current capability, meet Net Zero and deploy sustainable fuel and recycling concepts through the future. AFCP pioneered a uniquely collaborative approach to innovation. Uniting NNL’s unique nuclear infrastructure with the innovative skills and capabilities of more than 100 organisations across over 10 countries, AFCP delivered a suite of fuel cycle science themes with measurable impact across Britain’s evolving low-carbon landscape. Including delivering future roadmaps across advanced fuels and fuel cyclesd. Figure 1. Schematic representation of the UK fuel cycle with the focus areas of the Advanced Fuel Cycle Programme (AFCP) highlighted in green. Working collaboratively for a sustainable future Through bilateral and multi-lateral international relationships the UK continues to drive forward the importance of sustainable fuel cycles and the R&D needed to underpin these, such that options are available to decision and policy makers in a timely fashion. The fuel cycle underpins and will continue to be an essential part and enabler to the UK government’s net zero and national security objectives. c AFCP – Advanced Fuel Cycle Programme – Advancing fuel cycle innovation to secure a Net Zero future (nnl.co.uk) d https://afcp.nnl.co.uk/wp-content/uploads/sites/3/2021/06/AFCP-Advanced-Nuclear-Roadmaps.pdf https://afcp.nnl.co.uk/ https://afcp.nnl.co.uk/wp-content/uploads/sites/3/2021/06/AFCP-Advanced-Nuclear-Roadmaps.pdf 21 22 ACTINIDE AND FISSION PRODUCTS SEPARATION 23 Overview of the Material Recovery and Waste Form Development Program Kenneth C Marsden 12525 N Fremont Ave, Idaho Falls, ID 83402 The Material Recovery and Waste Form Development program is led by and funded through the Office of Materials and Chemical Technologies inside the U.S. Department of Energy, Office of Nuclear Energy. It is a research program with a mission to develop advanced fuel recycling technologies to improve resource utilization, reduce repository burden, limit proliferation risk and improve economics. Program activities include (1) use of recycling technologies to produce HALEU materials for advanced reactor fuel-fabrication R&D needs; (2) resolution of nuclear materials separation and recovery challenges for various advanced reactor fuel designs; (3) development of efficient and economical technologies for commercially viable future industrial deployment; and (4) expansion of capabilities and knowledge in nuclear chemistry for a broad range of nuclear applications. Technical areas inside the research program include simplified aqueous separations, vapor phase processes, pyrochemical processes, molten salt fuels, recycling off-gas management, and advanced waste form development. 24 25 Lab-Scale Pulsed Columns Trials for a New Nuclear Fuel Recycling Process Garzon Losik German, Lamadie Fabrice, Roussel Hervé CEA, DES, ISEC, DMRC, Univ. Montpellier, Marcoule, France Context and goals CEA is currently developing a new spent nuclear fuel reprocessing process, which is especially effective for fuels with high plutonium content. The process uses an N,N-dialkylamide extracting molecule. To validate the process chemistry, complete extraction cycles are performed using surrogate feed solutions (uranium and plutonium) or actual spent fuel solutions. These tests are performed in laboratory-scale mixer-settlers, which have the advantage of being miniaturisable and easy to operate in a nuclear confinement enclosure [1]. However, the pulsed column will be the device selected for a future facility, as in the current Orano plant at La Hague. This apparatus offers several advantages over the mixer-settler, such as cruds management, criticality control, and the absence of moving parts. It is mandatory to ensure that this type of apparatus performs correctly with the new solvent, both in terms of hydrodynamics and mass transfer efficiency. The knowledge and experience of CEA and Orano in the operation of pulsed columns with the current TBP-based solvent cannot be directly applied to the new solvent due to its higher viscosity when loaded with uranium. Therefore, a new study on the operation of pulsed columns is necessary. The latter involves both experimental work and a modeling and simulation approach to support scaling up to an industrial size. This communication presents the results of the experimental part of this approach. The objective of this experimental work is twofold: first, to demonstrate the suitability of the new solvent to be used in pulsed columns under the conditions specified for the new industrial process (including specific flow rates, flow rate ratios, and uranium load); and second, to collect data to validate the modelling of the pulsed column hydrodynamics. Initial tests were conducted using the smallest columns available at the CEA (15 mm in diameter), in order to minimize the required quantities of uranium and solvent. These tests allowed us not only to optimize the column packing, taking into account the solvent viscosity, but also to demonstrate the efficiency of these columns for uranium extraction and back- extraction. When scaling up from 15 mm columns, the first larger diameter chosen is 25 mm. This diameter allows for a significant increase in flow rates (by a factor of 2.8, corresponding to the cross-section ratio). Furthermore, it is the smallest diameter suitable for disc and doughnut packing, which is the type of packing used in industrial columns. Material The columns, of a height of 2 meters, used in the experiments are made of glass and have stainless steel packing with settlers located at both ends of them. A pneumatic type pulser, similar to those used at industrial scale at La Hague plant, is used to ensure mixing inside the column. Fluids are circulated using pumps combined with Coriolis flowmeters. Different types of characterisation are possible. - For single-phase tests: measuring residence time distribution by injecting an optical tracer on the inlet side of the column and monitor its concentration on the outlet side using an online refractometer. - For two-phase tests: measuring the dispersed phase hold-up by sampling the emulsion; additionally, capturing images using an endoscope (SOPAT GmbH) to obtain the droplet size distribution [2]. Single-phase tests The tests begin with a single phase, either aqueous (simple distilled water) or organic (a mixture of alkanes with the same viscosity as the N,N-dialkylamide based solvent loaded with uranium). 26 The first objective of these tests is to verify the possibility of obtaining a pulsating movement in the column, despite the solvent's viscosity. The main purpose is to determine the axial mixing by measuring the residence time distribution. Axial mixing is a critical parameter for column efficiency and is also a validation criterion for CFD simulations. It was found that axial mixing is not dependent on the flow rate or phase nature, but solely on the pulsation amplitude. A correlation derived from previous studies on other fluids correctly estimates this axial mixing [3]. Two-phase tests Various parameters have been tested including flow rates (ratio of organic to aqueous phase and total flow rate of both phases) and the pulsation amplitude. Both types of emulsions, oil-in-water and water-in-oil, were tested to cover the majority of operating conditions of the columns in the future process. Several hundred sets of operating parameters were tested. Among other results, the measurements allow: - To plot the range of correct column operation on a Sege and Woodfield diagram, - To monitor the evolution of the dispersed phase hold-up as a function of flow rates and pulsation amplitude, - To quantify the trends of the droplet size distribution as a function of these parameters. It appears that this distribution is particularly sensitive to pulsation, with size decreasing as amplitude increases. Conclusion and outlook It has been experimentally demonstrated that the N,N-dialkylamide based solvent enables proper hydrodynamic operation of pulsed columns with disc and doughnut packing. The study will continue by gradually approaching the operating conditions of industrial columns. The next steps will be to test extraction cycles with nitric acid and then with uranium in 25 mm diameter columns, and conducting hydrodynamic tests with uranium-loaded solvent in 100 mm diameter columns. Various aspects of the emulsion in a 25 mm diameter pulsed column with disc and doughnut packing References [1] Design, development and testing of miniature Liquid-Liquid Extraction contactors for R&D studies in nuclear environment, H. Roussel & S. Charton, Procedia Chemistry 21 (2016) 487-494 [2] Automated drop detection using image analysis for online particle size monitoring in multiphase systems, S. Maaß et al., Computers & Chemical Engineering 45 (2012) 27-37 [3] Axial dispersion in pulsed disk and doughnut columns: A unified law, S. Charton et al., Chemical Engineering Science 75 (2012) 468-477 27 Potential of Aggregation Control for Solvent Extraction Separation Cyril Micheau,a Yuki Ueda,a Ryuhei Motokawa,a Kazuhiro Akutsu-Suyama,b, c Norifumi L. Yamada,d Masako Yamada,d Sayed Ali Moussaoui,e Elizabeth Makombe,e Daniel Meyer,e Laurence Berthon,f Damien Bourgeoise a Materials Sciences Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan - b Comprehensive Research Organization for Science and Society, Tokai, Ibaraki 319-1106, Japan - c J-PARC Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan - d Institute of Materials Structure Science, High Energy Accelerator Research Organization (KEK), Tsukuba, Ibaraki 305-0801, Japan - e Institut de Chimie Séparative de Marcoule, ICSM, CEA, CNRS, ENSCM, Univ Montpellier, BP 17171, Marcoule, 30207 Bagnols-sur-Cèze, France - f CEA, DES, ISEC, DMRC, University of Montpellier, Marcoule, 30207 Bagnols-sur-Cèze, France Among the several processes developed for the advanced reprocessing of spent nuclear fuel, the DIAMEX (DIAMide EXtraction) process uses malonamide molecules for the separation and co-extraction of actinides and lanthanides from the raffinate of the PUREX (Plutonium Uranium Reduction EXtraction) process. Several studies were dedicated to the so called third-phase formation that occurs at high metal loading or in specific acid conditions in aliphatic solvents.[1] It was usually concluded that this deleterious phase arises from the formation of supramolecular aggregates mainly composed of the extractant molecules.[2] On the other hand, recent study on the use of malonamide molecules for the extraction and separation of Pd(II) and Nd(III) by solvent extraction has pointed out two main driving forces that could explain the selectivity: coordination and extractant aggregation.[3] It has thus been demonstrated that, Pd(II) only needs coordination to be extracted whereas Nd(III) needs, in addition, the aggregation of the extractant. From these observations it has been highlighted that extractant aggregation can be beneficial for solvent extraction process if it can be controlled to enhance separation whilst avoiding third-phase formation. In order to confirm a potential correlation between the separation efficiency and the extractant aggregation, and to determine the physicochemical parameters that triggers the supramolecular organization, a model malonamide molecule used in the DIAMEX process, i.e. DBMA (N,N’-Dimethyl-N,N’-dibutyl-2-tetradecylmalonamide), was investigated. This extractant was dissolved, at room temperature, in pure n-heptane, pure toluene, and mixtures of both solvents at different ratios, and contacted with different nitric acid solutions (0 M, 1 M, 3 M and 5 M HNO3) with and without the presence of two metal ions, i.e. Pd(II) and Nd(III). These two solvents possess different relative permittivity at room temperature (i.e., εheptane = 1.924, εtoluene = 2.392), and topology with an aromatic ring for toluene that could lead to additional π-interaction with the extractant [4, 5]. For each experimental condition, distribution ratios of Pd(II) (DPd) and Nd(III) (DNd), as well as separation factors (SPd/Nd = DPd/DNd) were determined (see Figure 1a and b). In parallel, small-angle neutron scattering (SANS) technique was used to determine the size and shape of the aggregates formed (see Figure 1c and d). From these results, it can be said that n- heptane promotes the formation of large assemblies (high initial scattering intensity, I0, and large gyration radius, RG), whereas the presence of toluene tends to reduce the size of the assemblies (low I0, small RG) which was directly correlated to extraction separation efficiency. In addition, these results confirmed that water and nitric acid molecules contributes to the extractant aggregate’s structure leading to larger aggregates, as the scattering intensity increases, when the nitric acid concentration increases. Finally, it can be concluded that the selectivity of the process can be control by the size of the aggregates which is triggered by the organic solvent properties. In the present study, solvent nature was modified using different mixtures to control the process efficiency. In addition, the potential of aggregation control for solvent extraction separation can be directly applied for the extraction of Pd from automotive catalysts or shredded electronic scraps. This strategy can also be extended to most of the extractant molecules involved in the spent nuclear fuel recycling strategy as they all possess an amphiphilic nature and, for some of them, their aggregation properties have already been characterized in the literature. 28 Figure 1: Distribution ratio (DM) and selectivity factor (SPd/Nd) of Pd(II) and Nd(III) (a, b), and SANS data of DBMA (c, d) as function of toluene/n-heptane volume ratio, pre-contacted with 1 M nitric acid solution (a, c) and 3 M nitric acid solution (b, d) (initial metal concentration: [Pd(II)]aq=100mg/L, [Nd(III)]aq=200mg/L). I0 represent the scattering intensity at low scattering wave vector q, and RG is the gyration radius of the scatterers. References: [1] B. F. Smith , K. V. Wilson , R. R. Gibson , M. M. Jones, G. D. Jarvinen, Sep. Sci. Technol., 32:1-4, 1997, 149-173, (DOI: 10.1080/01496399708003192 [2] F. Testard, T. Zemb, P. Bauduin, L. Berthon, Liquid/Liquid Extraction: A Colloidal Approach, Vol. 19 (Ed. B. A. Moyer), CRC Press, Boca Raton, 2010, pp. 381–428 [3] R. Poirot, X. Le Goff, O. Diat, D. Bourgeois, D. Meyer, ChemPhysChem Commun., 17, 2016, 2112-2117 (DOI: 10.1002/cphc.201600305) [4] A. A. Maryott, E. R. Smith, Table of Dielectric Constants of Pure Liquids, National Bureau of Standards Circular 514, USA, 10, 1951 [5] G. Ritzoulis, N. Papadopoulos, D. Jannakoudakis, J. Chem. Eng. Data, 31, 1986, 146-148 (DOI : 10.1021/je00044a004) 29 Evolution of Uranium Recovery: Past, Present, and Future Perspectives Santa Jansone Popova1, Jopaul Mathew1, Jeffrey Einkauf1, Alexander Ivanov 1, Ilja Popovs1, Connor Parker1, Peter Zalupski2, Travis Grimes2, Corey Pilgrim2 1Oak Ridge National Laboratory, Oak Ridge, TN 37932 2Idaho National Laboratory, Idaho Falls, ID 83415 Over the past eight decades, extensive research has been dedicated to the domains of uranium recovery, purification, reprocessing, and recycling, propelled by its critical applications in nuclear technology. Nowadays, the nuclear power stands as an important source of electricity, contributing approximately 10% to global electricity generation.1 The hydrometallurgical process for uranium extraction has undergone notable evolution throughout this period. The REDOX plant, commissioned in 1951, was the very first reprocessing facility based on countercurrent, continuous flow separation of plutonium and uranium utilizing aliphatic ketone as an extractant.2 A few years later, in 1955, the initiation of operations at the PUREX plant employed tributyl phosphate as an extractant.2,3 The exploration of organic, lipophilic extractants, including trialkyl phosphine oxides4 and monoamides (GANEX 1st cycle process)5, as well as lipophilic ion exchangers like protonated trialkylamines (AMEX process)6, has been integral to the ongoing pursuit of efficient U(VI) separation in solvent extraction methods7. These developments underscore the dynamic nature of uranium processing methodologies, emphasizing the continuous quest for enhanced technologies in the realm of nuclear fuel cycle management. This presentation will provide a concise overview of challenges encountered in established uranium recovery processes, encompassing issues like the stability of organic extractants under process-relevant conditions, solvent hydrodynamic properties, and variations in selectivity. A central focus will be on novel solvating extractants designed for U(VI) recovery. Particular attention will be given to exploring the impact of varying alkyl group size within extractants on both selectivity and stability. The evolution of uranium processing spanning more than 80 years, bridging historical achievements with the latest cutting-edge developments, providing valuable insights into the continual refinement of uranium recovery methodologies will be presented. References: [1] World Nuclear Association, World Nuclear Performance Report 2023. World Nuclear Association (2022). [2] M. S. Gerber, The Plutonium Production Story at the Hanford Site: Processes and Facilities History. Report WHC-MR-0532, Westinghouse Hanford Company (1996). [3] W. B. Lanham, T. C. Runion, PUREX Process for Plutonium and Uranium Recovery. Report ORNL-479, Oak Ridge National Laboratory (1949). [4] C. A. Horton, J. C. White, Separation of Uranium by Solvent Extraction with Tri-n-octylphosphine Oxide. Analytical Chemistry 1958, 1779–1784. 30 [5] T. H. Siddall, III, Effects of structure of N,N-disubstituted amides on their extraction of actinide and zirconium nitrates and of nitric acid. Journal of Physical Chemistry 1960, 1863‒1866. C. Musikas, Potentiality of Nonorganophosphorus Extractant in Chemical Separations of Actinides. Separation Science and Technology 1988, 1211‒1226. S. Suzuki, Y. Sasaki, T. Yaita, T. Kimura, Study on Selective Separation of Uranium by N,N-dialkylamide in ARTIST Process. ATALANTE 2004. M. Miguirditchian, L. Chareyre, C. Sorel, I. Bisel, P. Baron, M. Masson, development of the GANEX process for the reprocessing of Gen IV spent nuclear fuels. ATALANTE 2008. [6] D. J. Crouse, K. B. Brown, The Amex Process for Extracting Thorium Ores with Alkyl Amines. Journal of Industrial & Engineering Chemistry 1959, 1461–1464. [7] J. D. Law, Aqueous Reprocessing of Used Nuclear Fuel. Report INL/MIS-17-40915, Idaho National Laboratory (2018). C. R. Edwards, A. J. Oliver, Uranium Processing: A Review of Current Methods and Technology. JOM 2000, 12–20. 31 Experimental and Modeling Study of Uranium(VI) and Nitric Acid Extraction With a N,N-Dialkylamide Solvent Thibau Blanc1, Donatien Gomes Rodrigues1, Pauline Moeyaert1*, Thomas Dumas1, Philippe Guilbaud1* 1CEA, DES, ISEC, DMRC, Univ Montpellier, Marcoule, France * Corresponding Author, E-mail: pauline.moeyaert@cea.fr Innovative solvent extraction processes are currently under development at CEA for the reprocessing of MOX spent nuclear fuels. N,N-dialkylamides demonstrated their potentiality in the recovery and recycling of plutonium and uranium, as an alternative to TBP. First, they exhibit a good stability towards radiolysis and hydrolysis. Secondly, distribution ratios of plutonium(IV) and uranium(VI) with N,N-dialkylamide are such that their extraction and separation is possible, without using any reducing agent for the uranium – plutonium partitioning. The extraction process is modeled using thermodynamic equilibrium calculations. Those calculations are based on activity coefficient determination for both aqueous and organic phases and on solving mass action law. For these models, assumptions have to be made related to the speciation of the complexes (N,N-dialkylamides- metallic cation for instance). Therefore, the knowledge of this speciation is essential to ensure a good predictivity of the model. A bibliographical state of the art was established, and allowed to highlight the most promising methods described in the literature in order to obtain speciation diagrams of the complexes formed during the extraction of nitric acid and uranium(VI) with a N,N-dialkylamide solvent. Those methods include various spectroscopy studies, such as UV- Vis and IR as well as numeric processes to deconvolve the resulting spectra (including Principal Component Analysis). Batch experiments were performed to study the behavior of water, nitric acid and U(VI) towards their extraction by a N,N-dialkylamide and to determine extraction isotherms. All extraction experiments were performed at 25°C by contacting the N,N-dialkylamide diluted in IP185 isane with aqueous solutions at various nitric acid and uranium concentrations. All organic phases were also characterized thanks to spectroscopic technics to obtain information about the speciation of the complexes in the organic phases. Experimental distribution data were described with a physicochemical model based on the application of the mass action law on each extraction equilibrium and assumptions about the stoichiometry of the complexes formed in the organic phase. Deviations from ideality in aqueous phase were estimated by calculating the activity coefficient of each component according to the "simple solutions" model. Deviations from ideality were also determined in organic phase thanks to Sergievskii-Dannus approach. During the modeling work, assumptions have to be made regarding the speciation of the formed complexes. Agreement between simulated and experimental speciation allows refine the species hypothesis. That is why the experimental speciation data has been included as an input to the previously described model that was only fed with distribution data. The best group of complexes is the one that give the best agreement between experimental and calculated data for both distribution and speciation data. It is expected that this additional data will provide to the model a better robustness for various conditions (such as nitric acid, solvent and U(VI) concentrations). Acknowledgments The authors want to acknowledge Orano and EDF for financial support. 32 33 Feasibility Study on PUREX-NUMAP Hybrid Reprocessing: Precipitation-Based Recovery of U(VI) from Organic Phases with 30% TBP Ririka Tashiro1), Satoru Tsushima1)2), Koichiro Takao1) 1)Laboratory for Zero-Carbon Energy, Tokyo institute of technology 2-12-1-N1-32, O-okayama, Meguro-ku, 152-8550 Tokyo, Japan. 2)institute of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendrof, Bautzner Landstrasse 400, 01328 Dresden, Germany Email: tashiro.r.ab@m.titech.ac.jp For nuclear fuel recycling, the reprocessing process is the most essential. Currently, the PUREX method is exclusively employed in the spent fuel reprocessing, where U and Pu to be recycled are separated from fission products and minor actinides through selective solvent extraction using tributyl phosphate (TBP). This method is beneficial to gain high decontamination factor and technological reliability through its long history over 80 years. However, unavoidable concerns are still present in terms of potential isolation of Pu, safety risks arising from large excess amounts of organic solvents, and operation complexity through repeated extraction cycles. To improve the nuclear security aspect, Pu is once separated from UO2 2+ through reduction of extractable Pu4+ to unextractable Pu3+, followed by remixing with U. To address these issues, we have proposed an advanced reprocessing technology based on selective crystallization of nuclear fuel materials (NFM), NUclear fuel MAterials selective Precipitation (NUMAP) [1], where double-headed N-alkylated 2-pyrrolidone derivatives (DHNRP, Fig.1) are employed for selective recovery of NFM, U, Pu and even Th. In our NUMAP method, separation of each NFM completes in a respective single step precipitation, where any organic solvents are no longer required to be used. Its another striking advantage is impossibility of Pu isolation, while Pu is always present with U even after separation. Herein, we propose installation of the NUMAP principle to the U/Pu separation step after their primary decontamination in the PUREX process. If successful, there will be no chance for Pu isolation throughout the reprocessing process, making the security aspect of nuclear fuel cycle much improved. As a first step to know feasibility of this PUREX-NUMAP hybrid reprocessing, we studied behavior of U(VI) in 30% TBP/hydrocarbons after addition of DHNRP shown in Fig. 1 in terms of molecular and crystal structures of deposits as well as recovery yield of U(VI) from the organic phase. To simulate the primary decontamination in PUREX, aliquots of 3.0 M HNO3 aqueous solution containing 0.2 M U(VI) and 30% TBP/hydrocarbons (hexane, n-dodecane) were added to glass vials in 1:1 v/v, and shaken for 5 min. To the separated organic phase, DHNRP equivalent to U(VI) initially loaded was added as being done in NUMAP, followed by sonication. After centrifugation, the separated supernatant was mixed with 0.1 M HNO3(aq) in 1:1 v/v, and shaken for 5 min. The U concentrations in all aqueous phases were measured by ICP-AES to calculate efficiencies of respective steps in the PUREX-NUMAP hybrid reprocessing. The obtained precipitation was also characterized by pXRD and FT-IR. For crystal structure analysis, the organic phase (100 μL) after the primary decontamination was layered in a glas