Maria Skłodowska-Curie: Commemorating 100th anniversary of her Nobel Prize in Chemistry
NUTECH-2011
International Conference on Development and Applications of Nuclear Technologies
Book of Abstracts
11-14 September 2011 AGH Krakow
AGH University of Science and Technology Faculty of Physics and Applied Computer Science Krakow 2011 Technical editors: Joanna Dudała Zdzisław Stęgowski
Cover design: Joanna Dudała
Photograph: Beata Ostachowicz
This book contains original materials as submitted by authors, without any editorial amendments, except adaptation of the page layout.
Acknowledgements The Conference Organizing Committee would like to thank the Ministry of Science and Higher Education, Warsaw, Poland for financial support in printing the conference materials.
ISBN 978-83-925779-1-1
Copyright © by Faculty of Physics and Applied Computer Science, AGH Krakow, 2011 Organizers & Sponsors
AGH University of Science and Technology Faculty of Physics and Applied Computer Science
Institute of Nuclear Chemistry and Technology
lMCrt.ori'AEA g International Atomic Energy Agency
International Atomic Energy Agency «III CRA, jm IhmlT KRAKÓW
Municipality of Krakow
National Atomic Energy Agency
Polish Nuclear Society
European Nuclear Society
Honorary Patrons:
Antoni Tajduś, Rector
Stanislaw Staszic AGH University of Science and Technology, Kraków, Poland
FacultWojciecy ohf ŁużnyPhysic,s Deaand nApplie d Computer Science, AGH-UST, Kraków, Poland
Organizing Committee
Chairman: Marek Lankosz Members: Dariusz Wegrzynek, Andrzej Kreft, Zdzisław Stęgowski, Joanna Dudała, Marek Ciechanowski, Grażyna Zakrzewska-Trznadel, Wojciech Migdał, Wojciech Głuszewski, (Piotr Urbański'')
International Advisory Committee:
J. Thereska, Albania K. Różański, Poland W. Aparecido Parejo Calvo, Brazil S. Taczanowski, Poland L.G. de Andrade e Silva, Brazil A. Tajduś, Poland El-Sayed A.Hegazy, Egypt (P. Urbański^), Poland R.M. Yousri, Egypt M. Waligórski, Poland P. Berne, France A. Buzdugan, R. of Moldova M. Rossbach, Germany D. Axente, Romania K. Ziemons, Germany C.C. Ponta, Romania M. Haji-Saeid, IAEA A. Ali Basfar, Saudi Arabia A. Markowicz, IAEA N. Miljevic, Serbia A. Faucitano, Italy M. Al-Sheikhly, USA P. Fuochi, Italy M.R. Cleland, USA G. Spadaro, Italy A. Chmielewski, Poland E. Dziuk, Poland A. Hubalewska-Dydejczyk, Poland E. Iller, Poland M. Jeżabek, Poland J. Karczewski, Poland A. Kreft, Poland M. Lankosz, Poland J. Nawrocki, Poland J. Niewodniczański, Poland P. Olko, Poland J. Pluta, Poland J.M. Rosiak, Poland
The main aim of the NUTECH Conference series is to bring together scientists working on the development and application of nuclear technologies and to discuss further research fulfilling the needs of modern society. Scientists and engineers are always encouraged to join the conference to discuss the work and results of research and also to share their valuable experience, contributing to the enrichment of the international community. The Conference has a long tradition in Poland where it started in 1960s. The last conference, organized in 2008, has attained an international status. The NUTECH-2011 Conference starts with two plenary sessions, entirely devoted to modern nuclear technologies for power generation and applications in life and material sciences. The Conference creates a great opportunity for interchanging views and experience, it enables presentation of the state of the art and modern trends in the research and applications. The NUTECH-2011 Conference is a joint organizational effort of several institutions including the AGH University of Science and Technology, Krakow, Institute of Nuclear Chemistry and Technology, Warsaw, Polish Nuclear Society and PGE EJ 1 Sp. z o.o. and the IAEA. It is hoped that the Conference Book of Abstracts, which covers over 160 oral and poster presentations contributed by hundreds of authors distributed worldwide, will provide readers with information about possibilities and results of the applications of nuclear technologies in the modern economy of the XXI century.
Marek Lankosz and Dariusz Wegrzynek
Contens
CONTENS 11
NUCLEAR ENERGY IN POLAND 27
NUCLEAR ENERGY STRATEGY AND INTERNATIONAL COOPERATION OF JAPAN 28 S. Machi RADIOACTIVE WASTE MANAGEMENT APPROACH AND PRIORITIES IN FRANCE: A METHODOLOGY ACQUIRED ON THE BASIS OF LESSONS LEARNT 29 B. Faucher THE SAFETY INFRASTRUCTURE OF THE POLISH NUCLEAR ENERGY PROGRAMME - PRESENT AND FUTURE CONCERNS 30 M.P.R. Waligórski TRANSMUTATIONS OF ACTINIDES IN FUSION-DRIVEN SYSTEMS - A MISCONCEPTION OR EFFECTIVE SYNERGY ? 31 S. Taczanowski URANIUM DEPOSITS AND NUCLEAR WASTE REPOSITORY IN POLAND - AN OVERVIEV 32 J. Nawrocki
APPLICATIONS OF NUCLEAR TECHNIQUES 33
ELECTRON ACCELERATORS IN ENVIRONMENT PROTECTION TECHNOLOGIES 34 A.G. Chmielewski HANDHELD ED-XRF ANALYZERS AND THEIR ROLE IN INDUSTRY, SCIENCE AND IMPROVING THE QUALITY OF OUR ENVIRONMENT 35 S. Piorek HERITAGE OF MARIA SKŁODOW SKA-CURIE 36 J. Niewodniczański NEW PROBES FOR MOLECULAR IMAGING AND RADIOTHERAPY IN NUCLEAR MEDICINE - DIRECTIONS AND DEVELOPMENTS 37 R. Mikolajczak REVIEW OF THE IAEA ACTIVITIES TO SUPPORT APPLICATIONS OF NUCLEAR TECHNIQUES FOR CHARACTERIZATION AND PROTECTION OF CULTURAL HERITAGE OBJECTS 38 A. Markowicz, A. G. Karydas, R. Padilla-Alvarez
11 NUTECH-2011
NUCLEAR ENERGY AND MANAGEMENT OF RADIOACTIVE WASTES 39
ACCURATE AND RELIABLE RECONSTRUCTION OF THE ISOTOPE SPECIFIC ACTIVITY CONTENT IN STANDARDIZED NUCLEAR WASTE DRUMS BY SEGMENTED GAMMA SCANNING 40 T. Krings, E. Mauerhofer ANALYSIS OF URANIUM SUPPLY FROM DOMESTIC RESOURCES 41 G. Zakrzewska-Trznadel, K. Frąckiewicz, B. Zielińska, I. Herdzik-Koniecko, P. Biełuszka, A. Miśkiewicz, K. Szczygłów, S. Wołkowicz, R Strzelecki, K. Kiegiel CONCENTRATION OF LIQUID RADIOACTIVE WASTE USING BIOPOLIMER- ENHANCED ULTRAFILTRATION 42 G. Zakrzewska-Trznadel, A. Miskiewicz, L. Fuks, K. Kulisa DESIGN INNOVATIONS IN MANAGING a-C ONTAMINATED UNSERVICEABLE GLOVE BOXES 43 D.S. Sandhanshive1, S.R Shendge1, P.P. Mazumdar1, A. Kumar, M.N.B. Pillai DETERMINATION OF SCALING FACTORS FOR DIFFICULT TO MEASURE NUCLIDES IN SPENT RESINS FROM CERNAVODA NNP 44 R.Toma, I. Prisecaru, C. Dulama IMPLEMENTING PUBLIC PARTICIPATION APPROACHES IN RADIOACTIVE WASTE DISPOSAL 45 G. Zakrzewska- Trznadel METHODS OF IMMOBILIZATION OF RADIOACTIVE ELEMENTS IN SYNROC MATERIALS 46 T. Smoliński, A. Deptuła, A. G. Chmielewski PHOSPHORANE MINERALS AND PHOSPHORIC ACID AS POTENTIAL SOURCES OF URANIUM 47 G. Zakrzewska-Trznadel, A. Oszczak, P. Biełuszka PRELIMINARY PROPOSAL FOR RADIOACTIVE LIQUID WASTE MANAGEMENT IN A BRACHYTHERAPY SOURCES PRODUCTION LABORATORY 48 Carla D. Souza1a, Roberto Vicente , Maria Elisa C. M. Rostelato, Carlos A. Zeituni, Jõao A. Moura, Eduardo S. Moura, Fábio R Mattos, Anselmo Feher, Osvaldo L. Da Costa, Estanislau B. Vianna, Laércio de Carvalho, Dib Karan Jr. PROMPT GAMMA CHARACTERIZATION OF ACTINIDES 49 C. Genreith, M. Rossbach, E. Mauerhofer,T. Belgya REDEFINITION OF LARGE LOSS OF COOLANT ACCIDENT (LOCA) IN CONTEXT OF SEISMIC EVENT 50 K. Demjancuková STUDY OF THORIUM - URANIUM FUEL CYCLE 51 P. Kalbarczyk, H. Polkowska-Motrenko, E. Chajduk THE DEPLETION ANALYSIS OF HELIOS EXPERIMENT USING MCB CODE .. 52 M. Oettingen, E. D'Agata, Ch. Döderlein, K. Tućek, J. Cetnar
12 Contens
THE GOLDEN MEAN BETWEEN FUSION AND FISSION - THE FUSION HYBRID 53 G. Wójcik, S.Taczanowski THE HYBRID SYSTEM FOR LIQUID LOW-LEVEL RADIOACTIVE WASTE TREATMENT WITH APPLICATION OF MEMBRANE PROCESSES 54 G. Zakrzewska-Trznadel, M. Harasimowicz, A. Miśkiewicz, A. Jaworska UAE CIVIL NUCLEAR PROGRAMME 55 K. Szornel
RADIATION CHEMISTRY 57
ACUTE TOXICITY ASSESSMENT OF FLUOXETINE HYDROCHLORIDE (PROZAC®) WHEN SUBMITTED TO ELECTRON BEAM IRRADIATION 58 D.R.A. Santos, V.S.G. Garcia, A.C.S. Vilarrubia, S.I. Borrely APPLICATION OF RADIATION TREATMENT OF CELLULOSE PULPS FOR PREPARATION OF DERIVATIVES AND MICROCRISTASLLINE CELLULOSE 59 H. Stupińska, E. Iller, D. Wawro, Z. Zimek, D. Ciechańska, E. Kopania CHANGES IN PROPERTIES OF HYDROBIODEGRADABLE FILM BASED ON ALIPHATIC-AROMATIC COPOLYESTERS TREATED BY IONIZING RADIATION 60 H. Kubera, K. Melski, K. Assman, W. Głuszewski, Z. Zimek, N. Czaja-Jagielska DOSE SENSITIVITY ENHANCEMENT ON POLYMER GEL WITH SUSPENDED GOLD PARTICLES 61 L.C. Afonso, F. Schöfer, C. Hoeschen, L.V.E. Caldas EFFECT OF IONIZING IRRADIATION ON TILAPIA (OREOCHROMIS NILOTICUS) SKIN 62 C.A.P. Frose, E. Moura, R.B. Yamaguishi, E.S.R. Somessari, C.G. Silveira, E. Leme, A.B.C. Geraldo, J.E. Manzoli EFFECT OF TRANSITION METAL SALTS ON COLLOR, GLOSS AND HARDNESS OF EB-CURED PIGMENTED COATINGS FOR POLYMERIC SUBSTRATES 63 Marcelo Augusto Gonçalves Bardi, Mara de Mello Leite Munhoz, Luci Diva Brocardo Machado ELECTRON BEAM TREATMENT OF EXHAUST GAS WITH HIGH NOX CONCENTRATION 64 A.G. Chmielewski, A. Pawelec, J. Licki Y-RAY INDUCED VULCANIZATION OF RUBBER COMPOUNDS CONTAINING PRISTINE AND MODIFIED SILICA 65 D. Dondi, A. Lostritto, L. Conzatti, M. Castellano, A. Turturro, S. Bracco, M. Galimberti, A. Buttafava, A. Faucitano
13 NUTECH-2011
HIGH-TEMPERATURE OXIDATION RESISTANCE OF STAINLESS STEEL DOPED WITH YTTRIUM USING ION IMPLANTATION 66 M. Barlak, J. Piekoszewski, Z. Werner, B. Sartowska, L. Waliś, W. Starosta, J. Kierzek, K. Bocheńska, R. Heller, R. Wilhelm, A. Kolitsch, C. Pochrybniak, E. Kowalska HUMIDITY EFFECT ON RADIATION CHEMICAL DEGRADATION YIELD OF CHITOSAN 67 S. Benamer, M. Mahlous, D. Tahtat, A. Nacer Khodja, M. §en INFLUENCE OF IONISING AND UV RADIATION ON TEMPLATE DEPOSITED MICROSTRUCTURES OF SILVER HALOIDS 68 M. Buczkowski, B. Sartowska, W. Starosta KINETICS AND MECHANISMS OF RADIATION-INDUCED DEGRADATION OF HEXACHLOROCYCLOHEXANE IN WATER 69 Hasan M. Khan, Sanaullah Khan LIFETIME PREDICTION OF CABLES INSTALLED IN NUCLEAR POWER PLANTS BASED ON ANTIOXIDANT DECOMPOSITION IN INSULATIONS .... 70 J. Boguski, G. Przybytniak, K. Mirkowski, A. Bojanowska-Czajka MECHANICAL EVALUATION OF PVC FILMS MODIFIED BY ELECTRON BEAM IRRADIATION 71 J.R. Cardoso, E. Moura, E.S.R. Somessari, C.G. Silveira, H.A. Paes, C.A. Souza, J.E. Manzoli, A.B.C. Geraldo PRELIMINARY GAS CHROMATOGRAPHY/MASS SPECTROMETRY EVALUATION OF POLYCHLORINATED BIPHENYLS REMOVAL FROM WASTEWATERS BY GAMMA IRRADIATION 72 M. Virgolici, I. Dobrica, I.R. Stanculescu, A.V. Medvedovici, M.M. Manea, M. Alexandru, C.D. Negut, M. Cutrubinis, I.V. Moise, C.C. Ponta R&D LABORATORY WITH LINEAR ACCELERATOR FOR RADIATION PROCESSING 73 M. Fülöp, L. Harmatha, M. Ziska, M. Nemec, I. Benkovsky RADIATION GRAFTING OF ACRYLIC ACID ONTO CHITOSAN BEADS FOR METAL IONSORPTION 74 S. Benamer, M. Mahlous, D. Tahtat, A. Nacer-Khodja, M. Arabi, H. Lounici, N. Mameri RADIATION MODIFICATION OF ELASTOMERS 75 W. Głuszewski, Z.P. Zagórski, M. Rajkiewicz RADIATION MODIFICATION OF THE PHYSICOCHEMICAL AND FUNCTIONAL PROPERTIES OF THE POLYSACCHARIDE FILMS 76 K. Cieśla RADIATION SYNTHESIS OF PVA/CHITOSAN HYDROGEL FOR WOUND HEALING ENHANCEMENT 77 A. Nacer Khodja, M. Mahlous, D. Tahtat, S. Benamer, S. Larbiyoucef, H. Chader, L. Mouhoub, M. Sedgelmaci, N. Ammi, M.B. Mansouri, S.Mameri
14 Contens
RADIATION SYNTHESIS OF SILVER MICRO- AND NANOPARTICLES EMBEDDED IN COTTON FABRIC 78 D.Chmielewska, W. Starosta RADIOLYTIC DECOMPOSITION OF DICLOFENAC IN WATER BY GAMMA IRRADIATION 79 A. Bojanowska-Czajka, M. Trojanowicz, D. Solpan, G. Kciuk, G. Nałęcz- Jawecki SELECTION OF THE MATERIALS FOR RADIATION CROSS-LINKED CABLES 80 G. Przybytniak, Z. Zimek, A. Nowicki, K. Mirkowski, J. Boguski WASTEWATER TREATMENT WITH MOBILE E-BEAM PLANT 81 B. Han, J.K. Kim, Y. Kim, W.G. Kang, N. Zomme WETTABILITY OF CARBON AND SILICON CARBIDE CERAMICS INDUCED BY THEIR SURFACE ALLOYING WITH Ti, Zr AND Cu ELEMENTS USING HIGH INTENSITY PULSED PLASMA BEAMS 82 M. Barlak, J. Piekoszewski, Z. Werner, B. Sartowska, L. Waliś, W. Starosta, J. Kierzek, K. Bocheńska, R. Heller, A. Kolitsch, C. Pochrybniak, E. Kowalska
DOSIMETRY AND RADIATION PROTECTION 83
A 2-D THERMOLUMINESCENCE DETECTOR SYSTEM BASED ON LiF:Mg, Cu, P AND CaSO4:Dy FOILS FOR QUALITY ASSURANCE IN RADIATION DOSIMETRY 84 M. Kłosowski, R Kopeć, J. Gajewski, D. Kabat, K. Kisielewicz, P. Olko, M. Ptaszkiewicz, T. Nowak, M.P.R. Waligórski A COMPARISION OF METHODELOGY OF DOSE CALCULATION METHODS FOR ASYMMETRIC FIELDS IN NUCLEAR TECHNOLOGIES 85 S.H. Masoumi A NOVEL MONITOR OF NEUTRON - GAMMA RADIATION DOWN TO ENVIRONMENTAL LEVELS 86 S. Pszona ALANINE DOSIMETRY OF 60 MeV PROTON BEAM AT IFJ PAN - PRELIMINARY RESULTS 87 B. Michalec, G. Mierzwińska, U. Sowa, T. Nowak, J. Swakoń APPLICATION OF LiF:Mg,Cu,P (MCP-N) THERMOLUMINESCENT DETECTORS FOR EXPERIMENTAL VERIFICATION OF RADIAL DOSE DISTRIBUTION MODELS 88 W. Gieszczyk, P. Olko, P. Bilski, L. Grzanka, B. Obryk APPLICATION OF RETROSPECTIVE BIOLOGICAL DOSIMETRY WITH DICENTRICS AND FISH TECHNIQUES FOR ACCIDENTAL EXPOSURES TO RADIATION IN POLAND (1996-2010) 89 A. Cebulska-Wasilewska, J. Miszczyk, J. Swakoń
15 NUTECH-2011
CHARACTERISTICS AND PROPERTIES OF SOLID STATE DETECTORS IN A 60 MeV PROTON BEAM 90 U. Sowa, T. Nowak, B. Michalec, G. Mierzwińska, J. Swakoń, P. Olko COMBINED TL AND OSL READOUT OF LiF DETECTORS 91 B. Marczewska, P. Bilski, E. Mandowska, A. Mandowski COMPARISON OF APPLICATIONS OF GENE MUTATION ASSAY IN Trad-SH CELLS FOR MONITORING AMBIENT AIR GENOTOXICITY AFTER CHERNOBYL AND FUKUSHIMA NUCLEAR POWER PLANT ACCIDENTS 92 A. Panek, J. Miszczyk, A. Cebulska-Wasilewska DEDICATED COMPUTER SOFTWARE TO RADIATON DOSE OPTIMIZATION FOR STAFF PERFORMING NUCLEAR MEDICINE PROCEDURES 93 J. Kosek, K. Matusiak DEVELOPMENT OF CALIBRATION PROCEDURE FOR DOSE CALIBRATORS WITH USE OF THE EFFICIENCY CURVES 94 T. Dziel, A. Patocka, A.Muklanowicz ENERGY AND DOSE RESPONSE OF 2-D THERMOLUMINESCENT FOILS: TYPE LiF:Mg,Cu,P AND CaSO4 FOR RADIOLOGY PURPOSES 95 R. Kopeć, M. Klosowski EVALUATION OF ABSORBED DOSE DISTRIBUTION IN "WIERZCHOWSKA GÓRNA" LIMESTONE CAVE 96 B. Karabin, A. Jung MEASUREMENT AND CALCULATION OF EXISTING GAMMA DOSE IN ENVIRONMENT DURING THE DETERMINATION OF NITROGEN CONTENT OF EXPLOSIVE MATERIALS BY PGNAA METHOD USING MCNPX CODE 97 M.N. Nasrabadi, S. Omidi NANODOSIMETRY - A NEW TOOL FOR DESCRIPTION OF RADIATION ACTION ON THE NANOSTRUCTURES 98 S. Pszona and A. Bantsar ON THE QUALITY OF RADIATION PROTECTION IN SELECTED NUCLEAR MEDICINE DEPARTMENTS PERFORMING SCINTIGRAPHY AND PET-CT IN POLAND 99 R. Kopeć, M. Budzanowski, A. Budzyńska, R. Czepczyński, E. Dziuk, M. Dziuk, J. Sowiński, A. Wyszomirska, M.P.R. Waligórski RADIATION RISK CAUSED BY ENHANCED NATURAL RADIOACTIVITY .... 100 B. Michalik
ENVIRONMENTAL STUDIES 101
BIOACCUMULATION OF POLONIUM (210PO) AND URANIUM (234U, 238U) IN PLANTS AROUND PHOSPHOGYPSUM WASTE HEAP IN WIŚLINKA (NORTHERN POLAND) 102 A. Boryło, B. Skwarzec, G. Olszewski, D. Strumińska-Parulska
16 Contens
CURRENT STATUS OF RADON AND RADIUM ACTIVITY MEASUREMENTS IN WATER AT THE FEDERAL UNIVERSITY OF TECHNOLOGY (UTFPR, BRAZIL) 103 Janine Nicolosi Corrêa, Jaqueline Kappke, Sergei A. Paschuk, Hugo R. Schelin, Valeriy Denyak, Allan F. N. Perna, Marilson Reque EVALUATION OF THE ORIGIN OF SULFATE IN THE WATER SOURCE KLJUC, SERBIA 104 N. Miljevic, D. Boreli-Zdravkovic, G. Dusan, B. Mayer MEASURMENT OF THE RADON CONCENTRATION OF AIR SAMPLES IN THE SARI CITY 105 A. Rahimi MODELLING OF CALENDAR TIME SCALES FOR LAMINATED LAKE SEDIMENTS IN NORTHERN POLAND 106 N. Piotrowska, W. Tylmann, M. Kinder, D. Enters NATURAL RADIOACTIVITY IN GROUNDWATER 107 M. Dulinski, N.D. Chau, P. Jodlowski, J. Nowak, K. Rozanski, M. Sleziak, P. Wachniew PLUTONIUM SPECIATION IN THE SOUTHERN BALTIC SEA SEDIMENTS .... 108 D.I. Strumińska-Parulska, B. Skwarzec, M. Pawlukowska POLONIUM, URANIUM AND PLUTONIUM BIOACCUMULATION IN MARINE BIRDS 109 D.I. Strumińska-Parulska, B. Skwarzec, A. Boryło, J. Fabisiak RADIATION DEGRADATION OF OIL POLLUTED SOILS AND ITS COMPONENTS 110 D. Abbasova RADIOCARBON FOR NUCLEAR ENERGY 111 A. Pazdur, N. Piotrowska, K. Tudyka RADIOCHEMICAL ANALYSIS OF RADIUM 226Ra IN ENVIRONMENTAL SAMPLES USING ALPHA SPECTROMETRY 112 A. Boryło, B. Skwarzec, G. Olszewski, D. I. Strumińska-Parulska RADIOISOTOPE INVESTIGATIONS OF COMPOUND TWO PHASE FLOWS IN OPEN CHANNEL 113 M. Zych, L. Petryka, J. Kępiński, R. Hanus, T. Bujak, M. Sleziak, E. Puskarczyk SOURCE APORTIONMENT OF PARTICULATE MATTER (PM10) COLLECTED IN KRAKOW, POLAND 114 L. Samek, G. Fira SPATIAL DISTRIBUTION OF EQUIVALENT GAMMA DOSE RATE IN THE VICINITY OF MINE WATER SEDIMENTATION PONDS, UPPER SILESIAN COAL BASIN 115 M. Sleziak, M. Duliński TRACE ANALYSIS OF VOLCANIC ASH AND IT'S LEACHING DYNAMIC S ... 116 S. Landsberger, B. Canion, C. Jacques
17 NUTECH-2011
URANIUM AND RADIUM ISOTOPES IN GEOTHERMAL WATERS 117 Nguyen Dinh Chau, J. Nowak
RADIOTHERAPY 119
ACTIVATION OF LINEAR MEDICAL ACCELERATORS - AN OVERVIEW 120 K. Polaczek-Grelik, B. Karaczyn APPLICATION OF THE MONTE CARLO CALCULATIONS FOR A DESIGNING OF THE SIMPLE ENERGY MODULATOR IN THE PASSIVE BEAM-DELIVERY TECHNIQUE FOR THE 50 MEV - 70 MEV PROTON BEAMS 121 M. Grządziel, A. Konefał, W. Zipper AUTOMATION SYSTEM FOR QUALITY CONTROL IN THE MANUFACTURING OF IODINE-125 SEALED SOURCES USED IN BRACHYTHERAPY 122 S. L. Somessari, A. Feher, F.E. Sprenger, M.E.C.M. Rostellato, J.A. Moura, O.L. Costa, W.A. Parejo Calvo DESIGN AND PERFORMANCE OF A SYSTEM FOR TWO-DIMENSIONAL READOUT OF GAS ELECTRON MULTIPLIER DETECTORS FOR PROTON RANGE RADIOGRAPHY 123 W. Dąbrowski, T. Fiutowski, B. Mindur, P. Wiącek, A. Zielińska DETERMINATION OF ENERGY SPECTRA OF THERAPEUTIC X-RAY BEAMS FROM MEDICAL LINACS FOR VARIOUS IRRADIATION CONDITIONS 124 M. Bakoniak, A. Konefal FETAL DOSE EVALUATION IN BREAST RADIOTHERAPY USING SHIELDING AND PHYSICAL AND ENHANCED DYNAMIC WEDGES 125 D. Filipov, H.R. Schelin, D.S. Soboll IMPROVEMENTS IN THE QUALITY CONTROL OF IRIDIUM-192 WIRE USED IN BRACHYTHERAPY 126 Osvaldo L. Costa, Carlos A. Zeituni, Maria Elisa C. M. Rostelato, João A. Moura, Anselmo Feher, Eduardo S. Moura, Carla D. Souza, Samir L. Somessari MONOLITHIC APPLICATORS OF 125I AND 106Ru APPLIED IN EYE CANCER BRACHYTHERAPY 127 I. Cieszykowska, A. Piasecki, T. Janiak, M. Żółtowska ,T. Barcikowski, M. Mielcarski
RADIOMETRIC MEASUREMENTS 129
A NEW MONTE CARLO SUPPORT FOR THE INTERPRETATION OF THE GAMMA-GAMMA BOREHOLE GEOPHYSICAL TOOL RESPONSES IN CASE OF HETEROGENEOUS BOREHOLE VICINITY 130 U. Wiącek, T. Zorski, U.Woźnicka, D.Dworak ANALYSIS OF NATURALLY OCCURING RADIOACITVE MATERIAL USING NEUTRON ACTIVATION ANALAYSIS AND PASSIVE COMPTON SUPPRES SION GAMMA-RAY SPECTROMENTRY 131 S. Landsberger, G. George, D.Tamalis, J. Jean-Louis
18 Contens
APPLICATION OF SEALED RADIOACTIVE SOURCES TO MEASURE AND CONTROL FLOW IN INDUSTRIAL PROCESSES 132 L. Petryka, H.J.Pant, M. Zych, D. Kosman, M. Sleziak, R. Hanus DEVELOPMENT OF THE MECHANICAL SYSTEM ON A THIRD- GENERATION INDUSTRIAL COMPUTED TOMOGRAPHY SCANNER IN BRAZIL 133 Wilson A. Parejo Calvo, Carlos H. de Mesquita, Francisco E. Sprenger, Fabio E. da Costa, Pablo A. Vasques Salvador, Diego V. de Souza Carvalho, Margarida M. Hamada EXPERIMENTALLY AND NUMERICALLY PREDICTED RESIDENCE TIME DISTRIBUTION IN CHEMICAL REACTORS 134 Z. Stęgowski, L. Furman, C.P.K Dagadu. IMAGING TECHNIQUE FOR TROUBLESHOOTING OF INDUSTRIAL EQUIPMENT BY GAMMA-RAY ABSORPTION SCANS 135 M.I. Haraguchi, H.Y. Kim, F.E. Sprenger, W.A. Parejo Calvo LABORATORY AUTOMATIC MEASURING SYSTEM OF GAMMA SPECIMENS 136 P. Filipiak, A. Jakowiuk, J. Bartak, B. Machaj, P. Pieńkos, E. Kowalska RADIOTRACERS AS AN EFFECTIVE TOOL FOR MEMBRANE PROCESSES INVESTIGATION 137 A. Miskiewicz, G. Zakrzewska-Trznadel SPECTRAL ANALYZES OF LIQUID-GAS MIXTURE FLOW IN PIPES 138 L. Petryka, M. Zych, A. Sokulska, R Hanus ULTRA-LOW ENERGY X-RAY CALIBRATION SOURCE WITH THE X-RAY TUBE 139 P. Mazerewicz, W. Czarnacki, A. Gójska, M. Kisieliński, M. Słapa, M. Traczyk
RADIOPHARMACEUTICALS AND RADIOISOTOPE PRODUCTION 141
188W/188Re GEL GENERATOR 142 M. Konior, E. Iller DEVELOPMENT OF 177LU-PHYTATE COMPLEX FOR RADIOSYNOVECTOMY 143 Hassan Yousefnia, Amir R. Jalilian, Samaneh Zolghadri, Ali Bahrami-Samani, Mohammad Mazidi1, Mohammad Ghannadi Maragheh IRIDIUM-192 SEED DEVELOPMENT FOR OPHTHALMIC CANCER TREATMENT 144 M.E.C.M.Rostelato, F.R Mattos, C.A. Zeituni, C.D.Souza, J.A.Moura, E.S.Moura,, A.Feher, O.L.Costa, F. S. Peleias Jr, J.RO. Marques, R BelfortNeto NUCLEAR DATA FOR THE CYCLOTRON PRODUCTION OF IRON-55 VIA VARIOUS REACTIONS 145 M. Sadeghi, N. Soheibi, T. Kakavand
19 NUTECH-2011
NUCLEAR MODEL CALCULATION ON CHARGE PARTICLE INDUCED REACTION ON Ti TARGET AND TARGETRY FOR 48 V PRODUCTION 146 M. Sadeghi, T. Kakavand, Z. Ansari PREPARATION AND QUALITY CONTROL OF 166Ho-DTPA-ANTICD20 FOR RADIOIMMUNOTHERAPY 147 Samaneh Zolghadri, Amir R. Jalilian, Hassan Yousefnia, Ali Bahrami-Samani, Simindokht Shirvani-Arani, Mohammad Ghannadi-Maragheh PRODUCTION OF FISSION PRODUCT Mo-99 USING HIGH ENRICHED URANIUM PLATES IN RESEARCH REACTOR MARIA THERMAL- HYDRAULIC AND NEUTRONIC ANALYSIS AND TECHNOLOGY 148 J. Jaroszewicz, Z. Marcinkowska, W. Mieleszczenko, K. Pytel PRODUCTION OF IODINE-125 IN NUCLEAR REACTORS: ADVANTAGES AND DISADVANTAGES OF PRODUCTION IN BATCH OR CONTINUOUS PRODUCTION IN CRYOGENIC SYSTEM 149 C. A. Zeituni, M.E.C.M. Rostelato, K.Jae-Son, Jun S. Lee, O.L. Costa, João A. Moura, A. Feher, E. S. Moura, C.D. Souza, F.R. Mattos, F.S. Peleias Jr., Dib Karam Jr. PRODUCTION, QUALITY CONTROL AND BIOLOGICAL EVALUATION OF 153Sm-TTHMP AND 153Sm-PDTMP AS A POSSIBLE BONE PALLIATION AGENT 151 Z. Naseri, A.Reza Jalilian, A. Nemati Kharat, A. Bahrami-Samani, M. Ghannadi- Maragheh THE INFLUENCE OF PARAMETERS OF TARGET MATERIAL ACTIVATION IN A NUCLEAR REACTOR ON EFFICENCY AND QUALITY OF LUTETIUM Lu-177 PRODUCTION 152 Z. Tymiński, E. Kołakowska, M. Konior, D. Pawlak, A. Patocka
CHARACTERIZATION OF MATERIALS 153
APPLICATION OF NUCLEAR TECHNIQUES FOR MATERIALS SURFACE CHARACTERISATION: OWN INVESTIGATIONS EXAMPLES 154 B. Sartowska, J. Piekoszewski, L. Waliś, W. Starosta, M. Barlak, R Ratajczak, M. Kopcewicz APPLICATION OF XRF AND GC-SYSTEM FOR SOURCE IDENTIFICATION OF ARCHEOLOGICAL SAMPLES FOUNDED IN MENTASHTEPE OF AZERBAIJAN 155 D. Abbasova DETERMINATION OF URANIUM (VI) AND THORIUM (IV) IN TECHNOLOGICAL PHOSPHORIC ACID SOLUTIONS 156 Z. Samczyński DETERMINING THE CONTENT OF 10B IN BORIC ACID BY MEANS OF THE THERMAL NEUTRON ABSORPTION TECHNIQUE 157 A. Bolewski Jr, M. Ciechanowski, A. Kreft
20 Contens
EFFECTIVENESS AND LIMITATIONS OF QUANTITATIVE NEUTRON IMAGING CORRECTIONS FOR ACCURATE CHARACTERISATION OF POROUS MEDIA 158 M.J. Radebe, F.C. de Beer, R.B. Nshimirimana, J.J. Milczarek, I,M. Fijał-Kirejczyk, J. Żołądek-Nowak, G. Nothnagel INSTRUMENTAL NEUTRON ACTIVATION ANALYSIS (INAA) FOR STEEL ANALYSIS AND CERTIFICATION 159 E. Chajduk, B. Danko, H. Polkowska-Motrenko METHOD FOR DETECTION OF HIDDEN EXPLOSIVES AND OTHER ILLICIT MATERIALS WITH USE OF NANOSECOND NEUTRON PULSES FROM PLASMA FOCUS DEVICE - MONTE CARLO MODELLING OF THE SYSTEM . 160 U. Wiącek, R. Miklaszewski, V.A. Gribkov, B. Gabańska QUANTITATIVE CHARACTERIZATION OF STRATIFIED MATERIALS BY CONFOCAL 3D MICRO-BEAM X-RAY FLUORESCENCE SPECTROSCOPY .... 161 M. Czyzycki, D. Wegrzynek, P. Wrobel, M. Lankosz THERMAL NEUTRON RADIOGRAPHY STUDIES OF DRYING OF RECTANGULAR BLOCKS OF WET MORTAR 162 I.M. Fijal-Kirejczyk, J.J. Milczarek, F.C. de Beer, M.J. Radebe, J. Żołądek-Nowak
BIOMEDICAL STUDIES 163
BIOKINETICS AND RADIATION DOSIMETRY FOR [4-14C] -CHOLESTEROL IN HUMANS 164 L.A. Marcato, M.M. Hamada,C.H. Mesquita ELEMENTAL QUANTIFICATION OF THIN TISSUE SAMPLES IN SR-XRF TECHNIQUE USING EXTERNAL STANDARDS 165 M. Szczerbowska-Boruchowska EXOGENOUS OR ONTOGENETIC FACTORS INFLUENCE ON CELLULAR RADIOSENSITIVITY AND CANCER INCIDENCE IN THE POPULATION MONITORED WITH BIOMARKERS APPLIED FOR THE RETROSPECTIVE BIOLOGICAL DOSIMETRY 166 A. Cebulska-Wasilewska INVESTIGATION OF IRON AND ZINC OXIDATION STATE IN DIFFERENT GRADES OF HUMAN BRAIN GLIOMAS USING XAFS SPECTROSCOPY 167 K. Wolska, A. Wandzilak, M. Czyżycki, P. Wróbel, M. Szczerbowska-Boruchowska, D. Adamek, E. Radwańska, M. Lankosz MEASURING THE LEVELS OF SOME TRACE ELEMENTS IN THE BLOOD OF PATIENTS SUFFERING FROM MULTIPLE SCLEROSIS USING NEUTRON ACTIVATION ANALYSIS 168 M.N. Nasrabadi, D. Forghani POTENTIALITIES OF 109CD-BASED X-RAY FLUORESCENCE FOR IN VIVO LEAD CONTENT IN BONE 169 M.L. Carvalho, E.Da Silva,J.P. Marques, M.T. Lima, A. Pejović-Milić, D.R. Chettle
21 NUTECH-2011
RADIOEMBOLIZATION OF COLORECTAL CANCER HEPATIC METASTASES USING YTTRIUM-90 MICROSPHERES - PRELIMINARY REPORT 170 Z. Podgajny, P. Piasecki, N.Szaluś, P. Zięcina, J. Barzał, K. Brzozowski, G. Kamiński SPECT AND PET IMAGING OF YTTRIUM-90 171 A. Budzyńska, N. Szaluś, E. Dziuk, M. Dziuk, G. Kamiński, D. Pawlak SUSCEPTIBILITIES TO RADIATION OF LYMPHOCYTES FROM CANCER PATIENTS IN COMPARISON TO REFERENCE GROUPS BY CLASSIC AND MOLECULAR CYTOGENETICS 172 J. Miszczyk, A. Cebulska-Wasilewska, B. Dobrowolska, Z. Drąg, Z. Rudek, W. Jędrychowski, Z. Dobrowolski THE SINGLE CELLS RESPONSE TO THE PROTON MICROBEAM IRRADIATION 173 A. Wiecheć, J. Lekki, E.Lipiec, W.Polak, W.M.Kwiatek THE USE OF SR-FTIR MICROSPECTROSCOPY FOR THE PRELIMINARY BIOCHEMICAL STUDY OF THE RAT HIPPOCAMPAL FORMATION TISSUE IN CASE OF PILOCARPINE INDUCED EPILEPSY AND NEUROPROTECTION WITH FK-506 174 J. Dudała, K. Janeczko, Z. Setkowicz, D. Eichert, J. Chwiej
NUCLEAR TECHNIQUES IN PRESERVING CULTURAL HERITAGE 175
APPLICATION OF GC TO STUDY RADIOLYSIS OF CULTURAL HERITAGE ARTEFACTS 176 W. Głuszewski IDENTIFICATION AND CHARACTERIZATION OF USED PIGMENTS USED IN ICON PAINTINGS FROM ST DEMETRIUS ORTHODOX CHURCH IN KORYTNIKI BY INSTRUMENTAL NEUTRON ACTIVATION ANALYSIS AND SCANNING MICROSCOPY 177 E. Pańczyk, M. Pańczyk, D. Chmielewska-Smietanko, J. Giemza, J. Olszewska-Swietlik, L. Giro INAA AND OTHER ANALYTICAL TECHNIQUES IN CULTURAL HERITAGE- ELEMENTAL ANALYSIS OF METAL THREADS FROM SILK VELVET IN WILANÓW MUSEUM-PALACE 178 A. Skłodowska, H. Polkowska -Motrenko, B. Danko, J. Dudek, E. Chajduk INSTRUMENTAL NEUTRON ACTIVATION ANALYSIS AS A SOURCE OF INFORMATION CONCERNING THE ORIGIN OF "RUTHENIAN ALABASTER" 179 T. Sliwa, E. Pańczyk, PHYSICAL CHEMICAL ANALYSIS OF y-IRRADIATED WOODEN ARTEFACTS 180 M. Virgolici, I.R. Stanculescu, M.M. Manea, P. Bugheanu, C.D. Negut, C.C. Ponta1, I.V. Moise, M. Cutrubinis
22 Contens
RADIATION TREATMENT OF LIBRARY AND ARCHIVAL COLLECTIONS FOR MICROBIOLOGICAL DECONTAMINATION 181 D. Chmielewska, U. Gryczka, W. Migdał, W. Daszewski, A. Kuberka, M. Chyrczakowska X-RAY FLUORESCENCE SPECTROSCOPY STUDY ON THAI DECORATIVE GLASS 182 K. Won-in, S. Pongkrapan, P. Dararutana X-RAY TECHNIQUES IN THE INVESTIGATIONS OF A GOTHIC SCULPTURE „THE RISEN CHRIST' 183 A. Mikołajska, M. Walczak, Z.Kaszowska, M. Urbańczyk-Zawadzka, R. P. Banyś
APPLICATIONS OF NUCLEAR TECHNOLOGIES IN AGRICULTURE AND FOOD PROCESSING 185
ACTIVITY OF E-BEAM IRRADIATION IN THE CONTROL OF RHIZOCTONIA SOLANI 186 L.B. Orlikowski, U. Gryczka, M. Ptaszek, W. Migdał GAMMA IRRADIATION INFLUENCE ON THE STRUCTURE OF POTATO STARCH GELS STUDIED BY SEM 187 K. Cieśla, B. Sartowska, E. Królak, W. Głuszewski ISOTOPIC METHODS FOR FOOD AND BAVERAGE AUTHENTICITY CONTROL - TRENDS AND STANDARYZATION 188 R. Wierzchnicki STUDY OF STABLE ISOTOPE COMPOSITION OF CHOSEN FOODSTUFFS FROM THE POLISH MARKET 189 K. Malec-Czechowska, R. Wierzchnicki STUDY ON RADIATION INDUCED RADICALS GIVING RISE TO EPR SIGNALS EMPLOYED FOR THE DETECTION OF RADIATION TREATMENT IN SUGAR CONTAINING FOOD 190 G.P. Guzik, W. Stachowicz, J. Michalik
TECHNOLOGICAL DEVELOPMENTS 191
A COMPARATIVE STUDY ON tHE PERFORMANCE OF RADIATION DETECTORS FROM THE HgI2 CRYSTALS GROWN BY DIFFERENT TECHNIQUES 192 J.F.T. Martins, F.E. Costa, R.A. Santos, C.H. Mesquita, M.M. Hamada ACCREDITED LABORATORY FOR MEAESUREMENTS OF TECHNOLOGICAL DOSES (LMTD) 193 A. Korzeniowska - Sobczuk, K. Doner, M. Karlińska
23 NUTECH-2011
COMPARISON OF GEANT4 SIMULATIONS WITH EXPERIMENTAL PROTON ENERGY LOSS FOR SOME THICK ABSORBERS 194 O. Yevseyeva, J. T. de Assis, I. Ievsieieva, H.R. Schelin, I. Evseev, E. Milhoretto, F. da Silva Ahmann, S.A. Paschuk, V.V. Denyak, K. S. Díaz, J.M. Hormaza, R.T. Lopes COMPUTER SIMULATIONS AND IMAGE RECONSTRUCTION FOR A PROTON COMPUTED TOMOGRAPHY SYSTEM 195 E. Milhoretto, H. Schelin, J. Setti, V. Denyak, S. Paschuk, I. Evseev, F. Silva, J. de Assis, O. Yevseyeva, R. Lopes, U. Vinagre Filho DIAMOND DETECTOR FOR A SPECTROMETRIC MEASUREMENT OF "LOST ALPHA PARTICLES" 196 J. Dankowski, K. Drozdowicz, B. Gabańska, A. Igielski, A. Kurowski, B. Marczewska, T. Nowak, U. Woźnicka EVALUATING THE SHIELDING PARAMETERS FOR NEUTRON FLUENCE OF 252CF SOURCE USING MCNP4C CODE 197 M.N. Nasrabadi GAMMA AND NEUTRON MAZE EFFICIENCY ENHANCEMENT 198 N.M. Omi, O. Rodrigues Jr., W.A.P. Calvo, P.R. Rela GASEOUS DETECTORS IN CURRENT HIGH ENERGY PHISICS EXPERIMENTS 199 S. Koperny, T.Z. Kowalski INCRUSTATION OF a-PARTICLE EMITTERS IN THE SOURCE BACKING: INFLUENCE ON ACTIVITY MEASUREMENTS 200 A. Fernández Timón, M. Jurado Vargas INFLUENCE OF GROWTH PARAMETERS ON TERMOLUMINESCENT PROPERTIES OF CVD DIAMOND 201 M.Mitura-Nowak, A.Karczmarska, B. Marczewska, M. Perzanowski, M. Marszałek MATERIALS FOR RADIATION DETECTORS AND DOSIMETERS - THEMOLUMINESCENCE PROPERTIES OF RARE EARTH DOPED SCINTILLATING CRYSTALS 202 P. Bilski, A. Twardak, Y. Zorenko METAL-ORGANIC FRAMEWORK MATERIALS (MOF) AND THEIR APPLICATIONS 203 W. Starosta, B. Sartowska, A. Pawlukojć, L. Waliś, M. Buczkowski MODIFICATION OF THE STRUCTURE OF THE FILMS PREPARED BASING GAMMA IRRADIATED STARCH EXAMINED BY SCANNING ELECTRON MICROSCOPY 204 K. Cieśla, B. Sartowska NANOPORES WITH CONTROLLED PROFILES IN TRACK-ETCHED MEMBRANES 205 B. Sartowska, O.L. Orelovitch, A. Presz, P.Yu. Apel, I.V. Blonskaya PRECISION LOW POWER X-RAY GENERATOR 206 T. Kotowski
24 Contens
PREPARATION OF 57Co SOURCES FOR MÖSSBAUER SPECTROSCOPY 207 I. Cieszykowska, M. Żółtowska, M. Mielcarski, A. Piasecki, T. Janiak, T. Barcikowski PROMPT AND DELAYED GAMMA NEUTRON ACTIVATION ANALYSIS FOR THE ASSAY OF TOXIC ELEMENTS IN RADIOACTIVE WASTE PACKAGES ... 208 A. Havenith, J. Kettler, E. Mauerhofer SCATTERED NEUTRON COMPONENT IN DIGITAL THERMAL NEUTRON RADIOGRAPHS OF SIMPLE OBJECTS 209 J.J. Milczarek, F.C. de Beer, M.J. Radebe, I.M. Fijal-Kirejczyk, J. Żołądek-Nowak, A. Trzciński STUDY OF ANGULAR DISTRIBUTION OF NEUTRON EMITTED FROM PLASMAS USING NUCLEAR REACTIONS INDUCED IN INDIUM 210 S. Jednoróg, A.Szydłowski, M. Paduch, M. Scholz, B.Bieńkowska, R.Prokopowicz STUDY OF NUCLEAR LEVEL DENSITIES FOR EXOTIC NUCLEI 211 M.N. Nasrabadi, M. Sepiyani THE GROWTH AND SCINTILLATION CHARACTERISTICS OF LITHIUM DOPED CsI CRYSTALS 212 Maria da Conceição Costa Pereira, José Patrício Nahuel Cardenas, Tufic Madi Filho THE HPGE VIRTUAL POINT DETECTOR CONCEPT FOR RADIOACTIVE VOLUME-RING SOURCES BY MCNP4C SIMULATION 213 M.N. Nasrabadi THE PERFORMANCE OF STRAW TUBES 214 S. Koperny, T. Z. Kowalski VERY LOW COST MULTICHANNEL ANALYSER WITH SOME ADDITIONAL 215 FEATURES K. Tudyka, A. Bluszcz WIRELESS SYSTEM FOR RADIOMETRIC MEASUREMENTS 216 A. Jakowiuk, B. Machaj, P. Pieńkos, E. Kowalska, Paweł Filipiak, E. Swistowski MODELLING OF FUEL EQUILIBRIUM IN LEAD-COOLED FAST REACTORS .. 217 J. Cetnar, G. Domańska, P. Stanisz AUTHOR INDEX 219
25 NUTECH-2011
NUCLEAR ENERGY STRATEGY AND INTERNATIONAL COOPERATION OF JAPAN
Sueo Machi
Former Commissioner, Japan Atomic Energy Commission, (Fellow, Japan Atomic Energy Agency)
Japan's basic law of energy since 2002 states three basic policies are energy security, environmental compatibility and economic competitiveness. The energy security is the priority because indigenous energy of Japan is only 6%. Japan basic energy plan decided by the The Ministry of Economy, Trade and Industry (METI) in 2010 aims increasing nuclear and renewable power to 50% and 20%, respectively before 2030. In Japan there are 54 nuclear power plants of total 48.84GW capacity which have produced 29% of electricity in 2009, 2 under construction, and 12 under planning to be in operation by 2030. Cost of nuclear power is 5.2 Japanese Yen per kWh, which is most competitive in Japan. On 11th of March, 2011 the East Japan Great Earthquake attacked Fukushima Daiichi nuclear power station with tsunami of 14 meter high which caused unfortunate nuclear accident due to the damage of whole power supply including emergency generators. The Gov. of Japan and TEPCO has been making every effort to achieve the cold shutdown of 3 reactors before January 2012 which have damage of nuclear fuels due to loss of cooling. In the Ministerial Meeting of IAEA on Nuclear Safety from 20-24 June, 2011, the Gov. of Japan has reported the Fukushima nuclear accident and shared the lessons learned with international society. The Prime Minister of Japan stated at the OECD conference in May, 2011 that nuclear power remains to be a major energy source though renewable energy and saving energy should be more increased. Electricity from renewable energy should be 20% by early 2020's Japan has excellent record (OECD 2000) in CO2 emission per GDP, namely 0.21 ton/$1000, which is one tenth of that of China, thanks to energy saving and nuclear power. Emission of CO2 in Japan is 1.282 billion tons in 2008 which is 1.6 % higher than 1990 level and 4% of world emission. By the replacement of coal power plant of 1GW with nuclear power can save 6.40 million ton CO2 per year. In this regard nuclear power is essential to mitigate climate change. Japanese government has been contributing to international nuclear power community in both R/D and technology transfer to developing countries. The METI has trained more than 1000 nuclear power engineers and operators of developing countries in Asia and East Europe, while the Ministry of Education, Culture, Sports, Science and Technology (MEXT) has trained 1500 nuclear scientists and engineers of developing countries in Asia. The Cabinet Office and the MEXT are implementing a regional cooperation, the Forum for Nuclear Cooperation in Asia with the participation of 12 countries for promotion of nuclear technology applications including nuclear power.
28 Nuclear Energy in Poland
RADIOACTIVE WASTE MANAGEMENT APPROACH AND PRIORITIES IN FRANCE: A METHODOLOGY ACQUIRED ON THE BASIS OF LESSONS LEARNT
B. Faucher
Senior Expert at Andra International Division (French National Radioactive Waste Management Agency)
Properly addressing the issue of nuclear waste management is critical for an adequate acceptability of nuclear power and its sustainability. The matter is challenging on both its political fold and the technical one. In this respect, time is of the essence for a number of reasons: providing the population with enough and adapted information, explaining and discussing the options, developing a strategy and implementing it. Based on 40 years of lessons learnt and continuous improvement, a general framework has been set in France in order to address various issues such as "who does what ?", "who is responsible for what ?" and "which status for the various stakeholders ?", "waste quantities today ? and tomorrow ?". The main principles underlying this framework are: "nuclear safety first", "political decision-making & transparent process" and "waste producer pays". A clear and understandable waste classification with simple criteria applied to the waste inventory and production provides the basic data for waste management. Indeed, this radioactive waste issue is too often understood as and therefore reduced to spent fuel management. However, the first production of radioactive waste to be dealt with as soon as the reactor is commissioned will be the one of the so-called operational waste (mostly short- lived low- and intermediate-level waste - LILW), with approximately 100m3 per year and reactor (eg PWR type). It is therefore of paramount importance to deal with these LILW without delay, notably by organizing its safe management system over the long-term. Spent fuel or high-level related waste must be cooled down for at least a few decades and since volumes are quite small, they can be stored pending a final solution. In any case, the options for this final solution must be addressed as early as possible since it can impact further operational needs. At the end of the lifetime of the nuclear power plant starts the dismantling phase and a new stream of waste, mainly with a very low level of activity, will have to be disposed of. Finally, most countries do produce radioactive waste from "non-electronuclear" activities usually spread over the territory (medical, testing & analysis, food & water treatment, etc) and this waste also must be dealt with, for instance by taking stock of the existing organization and installations aimed at managing waste originated by the electronuclear industry. In conclusion, an agenda with waste volumes and management priorities can be drawn at the nuclear project inception and implemented consistently with it, as the technical long-term management solutions are available. This agenda will be also an important confidence- building instrument for the stakeholders and the public in general.
29 NUTECH-2011
THE SAFETY INFRASTRUCTURE OF THE POLISH NUCLEAR ENERGY PROGRAMME - PRESENT AND FUTURE CONCERNS
Michael P.R. Waligórski
Institute of Nuclear Physics, Polish Academy of Sciences, Radzikowskiego 152, 31 -342 Kraków, Poland and Centre Of Oncology Maria Sklodowska-Curie Memorial Institute Kraków Branch, Garncarska 11, 31-115 Kraków, Poland
According to the Resolution of the Council of Ministers of 13 January 2009 on the preparatory introduction of the Polish Nuclear Power Programme, the first nuclear power plant is scheduled to begin operation in 2020. This Resolution also indicated the role of the Government's Plenipotentiary for the Polish Nuclear Power Programme and designated the Polish Energy Group (PGE) as the Contractor (Operating Organisation) in the Polish Nuclear Power Programme. The "Framework time schedule for nuclear power activities" adopted by the Council of Ministers on 11 August 2009 stipulated that drafts of legal acts, necessary to establish and operate the nuclear energy sector should have been prepared by 31 December 2010. Appendix 3, "Energy Policy of Poland until 2030" sets out, among other matters, that the President of the National Atomic Energy Agency (PAA) should prepare drafts of legal acts on nuclear safety and radiological protection, ensuring implementation of the Polish nuclear power programme, by the end of 2010 (Measure 3.2, item 2). According to §2 Section 2.3 of the Regulation of the Council of Ministers of 12 May 2009 concerning the Government's Plenipotentiary for the Polish Nuclear Power Programme, the tasks of the Government's Plenipotentiary are to initiate and act to prepare drafts of legal acts necessary for the Programme's implementation. In the Agreement between the Government's Plenipotentiary and the Agency's President of March 23, 2010, the leading role of the Agency's President as the central body of the government administration competent over issues of nuclear safety and radiological protection in the elaboration of drafts of legal acts regulating issues of nuclear safety and radiological protection for nuclear facilities, had been emphasised. It should also be stressed that among the duties of the Government's Plenipotentiary is to elaborate drafts of legal acts that are necessary for implementing the Polish Nuclear Power Programme in other areas than nuclear safety and radiological protection, including regulations which form the decision- making structures related to the development of the nuclear energy sector in Poland and the investment processes for nuclear energy facilities, including promotion of such investments in the local communities. Another important area requiring attention of the Government's Plenipotentiary is the technical support organisation (TSO) for this Programme. The current developments of the Polish Nuclear Power Programme with regard to nuclear safety aspects will be reviewed against IAEA safety standards for protecting people and the environment, Safety Guide DS 424 "Establishing a Safety Infrastructure for a National Nuclear Power Programme". Based on recommendations of this document, suggestions as to future developments in this Programme will be made.
30 Nuclear Energy in Poland
TRANSMUTATIONS OF ACTINIDES IN FUSION-DRIVEN SYSTEMS - A MISCONCEPTION OR EFFECTIVE SYNERGY?
S. Taczanowski
Chair of Nuclear Energy, Faculty of Energy and Fuels, AGH University of Science and Technology, Cracow, Poland
The alliance of fusion with fission is a cause worthy of great efforts, as being able to ease (if not even to solve) the serious problems of both these forms of nuclear energy. 1) Fission. The most troublesome component of the High Level Waste (HLW) i.e. Minor Actinides (MA) shows disadvantageous physical properties: intense radioactivity, minute fraction of delayed neutrons, intense heat release, positive reactivity coefficients etc. Such features so much affect safety that exclude incineration of the MAs in critical systems. As a remedy - the MAs abatement in safer subcritical systems has been proposed, namely, in Fusion-Driven Actinide Incinerators (FDI) [1]. Besides the 14MeV neutron component may better incinerate all hardly fissionable actinides. Though the fusion technology is not yet ready, the problem of actinide waste seems last until the fusion hybrid systems are available. 2) Fusion. Very high investment costs caused by tokamak sizes and difficult technology put in doubt whether alone the minute demand for fuel (Li) and lack of danger of uncontrolled super-criticality prove sufficient for making it economically competitive. For illustration, the material consumption of tokamaks in some design (PPCS-AB) is exceeding 120 000 tons, while the mass of Li-Pb eutectic (tritium breeding material) amounts to 35 000 tons [2]. Tritium, due to its high solubility in most metals thus resulting in inventory attaining 10 kg [3] is a problem too. All the above problems may be solved with synergic union of fission with fusion. The performed preliminary evaluations demonstrated that a radical shift of energy production i.e. the energy gain from plasma to fissionable blanket is feasible. A reduction of the fusion component to about 2% at given power of the system, brings a radical drop in the value of plasma Q down to the level ~0.3 achievable in small systems (e.g. mirror devices - Gas Dynamic Traps - GDT). The sizes of the latter are rather comparable to those of fission reactors whose life cycle mass of contaminated steel amounts to a few thousand tons. In an FDI all the radiations from the plasma: corpuscular (i.e. neutrons, alphas and other ions) and electromagnetic ones are drastically reduced. Thus, proportionally - the radiation damage: its plasma- wall component and the neutron induced ones: gas production, DPA and to a degree - transmutations. Finally, last but not least, the fundamental safety of the system has been proved by simulation of its collapse that has shown preserving its subcriticality in this extreme state. In spite of all the above encouraging remarks a number of questions still await solution. E.g. a trade-off dilemma: system size vs. reduction of radiation damage, while the latter is quite unevenly distributed, is unsolved. From the point of view of fission the efficiency of MAs incineration requires further investigations, as well as the safety properties of the system. Summarizing, the concept of Fusion-Driven Actinide Incinerator - small, simple and cheaper deserves consideration also as an intermediate step towards the Pure Fusion that should bring near the development of Fusion Energy.
1. Taczanowski S (2009) Chapter 11 in: Advanced reactor technology options for utilization and transmutation of actinides in spent nuclear fuel, IAEA-TECDOC-1626, 219-237 2. Pampin R, Massaut V, Taylor NP, Revision of the inventory and recycling scenario of active material in near term PPCS models, Nucl. Fusion, 47, (2007) 469-476 3. Abdou M (2007) Overview of the Principles and Challenges of Fusion Nuclear Technology, http://www.fusion.ucla.edu/abdou/
31 NUTECH-2011
URANIUM DEPOSITS AND NUCLEAR WASTE REPOSITORY IN POLAND - AN OVERVIEV
Jerzy Nawrocki
jerzy.nawrocki@pgi. gov.pl
Polish Geological Institute - National Research Institute, Rakowiecka 4, 00-975 Warszawa, Poland
Anticipated economic and subeconomic resources of uranium deposits in Poland are rather of subordinate importance in term of nuclear fuel supply for our domestic nuclear power plants. However, they could play a significant role in the case when any political crisis will break deliveries of nuclear fuel from the abroad. Several sites with minor uranium bodies were documented in the Sudetes. They contain c.a. 2000 t of uranium and therefore has been classified as being out of economic value. More rich in uranium deposits geological structures occur in the eastern and northern parts of Poland. Up to 90 000 t of uranium is predicted in the Ordovician shale of the Podlasie Depression. However, in these deposit the mean content of uranium is rather low (70 g/t). Additionally, the uranium forms here a very complex metal-organic compound that cause difficulties in technology of its extraction. The most promising uranium mineralization was found in the Peri-Baltic Syncline. It was subsequently documented between 1975 and 1983. Irregular ore bodies with quite high content of uranium occurs here inside the brittle Triassic sandstones. However, a very strong dispersion and irregularity of occurrence was observed in studied drill cores. Moreover, the zone with uranium mineralization is not so shallow (c.a. 1 km of borehole depth). Because of this it has been postulated that any high-resolution geophysical method like 3D seismic image should be tested for their prospection. The best geological conditions and the selection of the best geological structures of Poland for a deep nuclear waste repository has been discussed for many years. They should be hermetic during quite long time, even up to 100 ka. That means its resistance to the water infiltration, tectonic movements and physical/chemical erosion. Previously our attention was focused mainly on the salt structures. Its richness, especially in the central and western parts of Poland and examples of acting repositories in such structures (Germany) could justify that point of view. On the other hand, however, recent studies of salt diapires disclosed their quite strong tectonic mobility. Therefore our attention should be rather addressed to the alkaline magmatic bodies of NE Poland. Their safe depth (~1 km in the Suwałki area), good physical properties and chemical composition, and safe tectonic location (East European Craton) allow to define them as the most perspective.
32 NUTECH-2011
ELECTRON ACCELERATORS IN ENVIRONMENT PROTECTION TECHNOLOGIES
Andrzej G. Chmielewski
Institute of Nuclear Chemistry and Technology, Warsaw, Poland Warsaw University of Technology, Poland
The radiation processing is well developed technology using the ionizing radiation in material treatment. As a industrial emitters of radiation gamma isotope sources and electron accelerators (at some high power units e/X converters are applied to assure deep penetration of radiation in high density materials) are used. The main fields of applications are sterilization of medical products, food hygenization and polymers and rubber modification. All over the world ca. 200 gamma irradiators and almost 2000 electron accelerators are in service. The very important field of applications is environment protection [1,2].These applications are based on high power accelerators application. The electron beam flue gas treatment has been implemented in the industry a few years ago. The first applications were devoted to the SO 2 and NO x treatment from coal fired boilers. The biggest ever built installation has been constructed in EPS Pomorzany, Poland. However the recent research results have demonstrated possibility of this technology application for volatile organic pollutants treatment [3] and mercury removal [4] as well. The other possible application of the technology is flue gas treatment from high sulfur oil fired boiler [5].The open field of applications is wastewater treatment, which has been illustrated at industrial scale on South Korea and sludge treatment for its hygenization or biogas production enhancement [1].The different fields of the particle accelerators applications has been reviewed and elaborated at DOE report "Accelerators for America's Future"[7], These all developments are discussed in the paper.
1. IAEA (2007) Radiation Processing: Environmental Applications, IAEA - Non serial publications, Vienna 2. Chmielewski AG (2005) Application of ionizing radiation to environment protection. Nukleonika 50;S3:S17 - S24 3. Sun YX; Chmielewski AG, Bulka S, et al. (2006) Influence of base gas mixture on decomposition of 1,4-dichlorobenzene in an electron beam generated plasma reactor, Plasma Chemistry and Plasma Processing 26; 4: 347-359 4. Jo-Chun Kim, Ki-Hyun Kim, Al Armendariz, Al-Sheikhly M (2010 May) Electron Beam Irradiation for Mercury Oxidation and Mercury Emissions Control, Journal of Environmental Engineering; 554 - 559 5. Basfar, AA; Fageeha OI; Kunnummal N,Chmielewski AG, Pawelec A, Licki J, Zimek Z (2008) Electron beam flue gas treatment (EBFGT) technology for simultaneous removal of SO2 and NOx from combustion of liquid fuels, Fuel 87; 8-9: 1446-1452 6. Sampa MH, Takacs E, Gehringer P, Rela PR Ramirez T, Amro H, Trojanowicz M, Botelho ML, Han B, Solpan D, Cooper WJ, Emmi SS, WojnarovitsE (2007) Remediation ofpolluted waters and wastewater by radiation processing. Nukleonika 52;4:137-144 7. DOE (2010) "Accelerators for America's Future", US DOE Washington, www.acceleratorsamerica.org/report/index.html
34 Applications of Nuclear Techniques
HANDHELD ED-XRF ANALYZERS AND THEIR ROLE IN INDUSTRY, SCIENCE AND IMPROVING THE QUALITY OF OUR ENVIRONMENT
S. Piorek
Thermo Niton Analyzers, LLC, 900 Middlesex Turnpike, Billerica, MA 01821
The first truly handheld, one-piece, isotope based EDXRF analyzer was introduced in 1995, followed in seven years by the first ever handheld with miniature x-ray tube (Fig. 1). These developments epitomize decades of developments in electronics, microprocessors, "room" temperature silicon detectors and miniature x-ray tubes. Combined with the latest Li+ battery technologies, they made possible design of a handheld ED-XRF analyzer weighing less than 1.5 kG, yet in terms of analytical capabilities, rivaling in most applications laboratory or bench-top version. The ruggedness and reliability of construction required of field-deployable instrument are likewise expected of the analytical and user interface software. The epitome of the XRF analysis - the fundamental parameters - is now a common feature on a handheld system and its practical implementation does not require the operator to hold a higher academic degree.
On average a new generation every 8 years MXI / ? 9> / m
Sir* IF piet« / X-ray Tubr łundTwitó ^ / - S Holds / Ocatcs and holds Calibration Calibration / /
Time, [years]
Figure 1. Close to 40 thousands of handheld XRF analyzers alone have been placed on the world market. Majority of them are used for analysis and identification of alloys by metal producing, recycling and fabrication industries, followed by risk critical industries such as aviation, oil refining and nuclear power. The second most popular application of the analyzers is soil screening for metallic contaminants. However, over the last decade handheld XRF analyzers became a routine tool for screening consumer electronic products for compliance with the RoHS and WEEE Directives and most recently for screening toys and consumer products for toxic elements such as lead, cadmium, mercury, etc. restricted by state or federal regulations. Speed of analysis, its nondestructive character and economics of use, are those features that made handheld XRF analyzer a tool of choice for industry, inspection and enforcing agencies. In this paper we briefly review the evolution of field portable XRF instrumentation and focus on its current state-of-art. We will discuss typical applications and performance of these instruments and will also address the challenges. Finally, we will conclude with look into future development trends one may expect of this versatile and powerful technique of EDXRF in its handheld and portable embodiment.
35 NUTECH-2011
HERITAGE OF MARIA SKŁODOWSKA-CURIE
J. Niewodniczański
niewodniczański@gmail. com
Faculty of Energy and Fuels, AGH University of Science and Technology, Cracow, Poland This year we pay a tribute to Maria Skłodowska-Curie on the occasion of the 100th anniversary of the Nobel Prize in Chemistry awarded to her for the discovery of polonium and radium. -Curie, through her masterly conducted research and quantitative characterization of faint traces of new elements, opened a pathway to applications of tracer atoms method in investigation of chemical reactions and industrial processes. Her lectures and papers inspired physicists to study atomic structure and atomic nucleus, as well as suggested a quantum nature of the subatomic world and a possible practical use of atomic energy. The availability of radium, she has discovered and managed to extract from uranium ore, opened new possibilities of curing patients suffering from cancer. To obtain a broader use of the "curietherapy" she had to create something new and unknown beforehand - a cooperation between scientists, politicians, and industry. Yet another new idea introduced by her to the scientific research was conception of international scientific teams she used to organize in her laboratory. Since then progress in physics, especially in the nuclear field, has been usually connected with collective efforts of young researchers. Thanks to her leadership and atmosphere she created in the laboratory numerous pupils of Madame Curie have achieved a broad recognition including awards of the Nobel Prize. -Curie to advancing the science in Poland, and especially to the development of Polish radiation and nuclear laboratories, cannot be overestimated. She founded centers of research and radiotherapy, donated instrumentation and samples of radium, initiated new programmes, but first of all - she trained numerous Polish scientists in her laboratory. -Curie in science, to discuss practical implications of her discoveries, and to recognize her impact on the organization of research, especially in Poland.
36 Applications of Nuclear Techniques
NEW PROBES FOR MOLECULAR IMAGING AND RADIOTHERAPY IN NUCLEAR MEDICINE - DIRECTIONS AND DEVELOPMENTS
R. Mikołajczak
Radioisotope Centre of IEA POLATOM, 05-400 Swierk-Otwock, Poland
In recent years the rapid expansion in the use of radionuclides for nuclear medicine has been observed involving various biologically active molecules as carriers. The potential usefulness of a particular radionuclide depends on many factors: physical data (half-life, energy of beta- particles, gamma ray emissions), production method including separation, labeling and targeting properties of a radionuclide carrier molecule. Several beta-emitting radionuclides for targeted radionuclide therapy can be produced in the nuclear reactors, including generator systems. The cyclotrons and radionuclide generators can provide a number of others including gamma or positron emitters, which can be used for diagnostics. The list of potentially useful isotopes is not yet closed. The knowledge of the potential targets which are present in the human body in sufficient quantities is also expanding. To provide an efficient radiopharmaceutical the identified target should be expressed in the tumor with high density and incidence, the radioligand should have a good access to it and its expression in normal tissues should be low. Somatostatin receptors have been shown as the target which could be effectively utilized for diagnostics with 99mTc or 68Ga labeled analogs or for therapy with beta-emitters such as carrier-free 90Y and carrier- added 177Lu. Since then several other potential targets have been identified, which are either expressed on the cell membrane (extracellular) such as transporters, neurotransmitter receptors, hormone receptors, neuropeptide receptors, growth factor receptors and tumour- associated antibody epitopes or intracellular like metabolic pathways, DNA/RNA or other organelles. The availability of radionuclides and potential targets allows to construct tailor- made radiopharmaceuticals for radionuclide diagnostics and therapy choosing from a variety of half-lives, energies and applicable chemistry. Radionuclides such as 131I, 90Y, 186/188Re, 166Ho and 153Sm have found applications in clinical procedures and have been used for cancer therapy, bone pain palliation, radiosynovectomy, intravascular radiation therapy and other disorders. Other radionuclides including 177Lu, 161Tb, 67Cu, 47Sc with promising physical and chemical properties still need to be explored. Approaches using alpha therapies may allow better targeting of residual disease in specific therapeutic frameworks. New developments in PET radiopharmaceuticals radiolabelled with 18F, 68Ga or Zr and the combination of PET for dosimetric applications involving the therapeutic matched radionuclide pairs such as 64/67Cu or 44/47Sc allow promising improvements in calculation of absorbed doses for an individual patient. Various techniques have been developed in order to improve and personalize the therapeutic effect of internal radiotherapy such as the radionuclide cocktail approach, loco-regional administration, pre-targeting or combination with chemotherapy, providing new therapy options. All these developments can be considered when designing new radiopharmaceuticals for personalized radionuclide diagnostics and internal therapy. The optimization of the radiopharmaceutical is crucial for its diagnostic or therapeutic efficacy and safety to the patients. Regulatory aspects: approval of clinical trial, Marketing Authorisation, GMP and Good Radiopharmacy Practice are additional challenge.
37 NUTECH-2011
REVIEW OF THE IAEA ACTIVITIES TO SUPPORT APPLICATIONS OF NUCLEAR TECHNIQUES FOR CHARACTERIZATION AND PROTECTION OF CULTURAL HERITAGE OBJECTS
A. Markowicz, A. G. Karydas, R. Padilla-Alvarez
Nuclear Spectrometry and Applications Laboratory, Physics Section, IAEA Laboratories, A-2444 Seibersdorf, Austria
The IAEA provides an extensive support to the laboratories in Member States in the field of applications of nuclear techniques for characterization, study and protection of cultural heritage (CH) objects. The nuclear techniques include analytical methods such as XRF, IBA and NAA, dating techniques and gamma- and electron-irradiation methods. Various modalities of the IAEA support will be outlined including Technical Cooperation (TC) projects used for establishing basic infrastructure and development of skilled personnel, Coordinated Research Projects (CRPs) used for coordination of research and development carried out by the laboratories in both the developed and developing countries, and applied/adaptive research and development activities carried out in the IAEA Laboratories at Seibersdorf. Emphasis will be on the recent results and selected applications of X-ray fluorescence (XRF) techniques including laboratory and in-situ measurements. XRF techniques provide information on the elemental composition of the CH objects which can be used in support of conservation and restoration, determination of production technology and origin, evaluation of alteration processes, diagnosis of previous modifications, and for preventive conservation. Major features of these techniques such as simplicity, speed of operation, immediate generation of multi-element analytical data, non-destructive character of analysis, portability and flexibility in terms of analysis of various objects made XRF techniques extremely attractive and fully recognized by the CH community. The facilities of the Nuclear Spectrometry and Applications Laboratory applied in support of CH field, include various laboratory and (trans)portable EDXRF spectrometers [1], micro-beam XRF and absorption techniques integrated in one instrument [2], SEM as well as access to IBA techniques and synchrotron radiation sources in the laboratories in Croatia and Germany. Most of the applications of the XRF techniques were performed in cooperation with the Museum of Fine Arts in Vienna for characterization of paintings, coins, sculptures, bronzes etc. The results of the applications for the elemental compositional characterization and 3D imaging will be presented.
1. Uhlir K, Griesser M, Buzanich G, Wobrauschek P, Streli C, Wegrzynek D, Markowicz A, Chinea-Cano E (2008) Applications of a new portable (micro) XRF instrument having low-Z elements determination capability in the field of works of art. X-Ray Spectrometry 37; 450-457 2. Wegrzynek D, Markowicz A, Bamford S, Chinea-Cano E, Bogovac M (2005) Micro-beam X-ray fluorescence and absorption imaging techniques at the IAEA Laboratories. Nuclear Instruments and Methods in Physics Research B 231; 176-182
38 NUTECH-2011
ACCURATE AND RELIABLE RECONSTRUCTION OF THE ISOTOPE SPECIFIC ACTIVITY CONTENT IN STANDARDIZED NUCLEAR WASTE DRUMS BY SEGMENTED GAMMA SCANNING
T. Krings, E. Mauerhofer
Institute of Energy and Climate Research - Nuclear Waste Management, Forschungszentrum Jülich GmbH,
Radioactive waste must be characterized in order to verify its conformance with the national regulations for intermediate and final storage. Segmented gamma scanning (SGS) is the most widely applied non destructive analytical technique for the characterization of radioactive waste drums. The isotope specific activity content is generally calculated assuming a homogeneous matrix and activity distribution for each measured drum segment. However, real radioactive waste drums exhibit non-uniform isotope and density distributions most affecting the reliability and accuracy of activities reconstruction in SGS. The presence of internal shielding structures in the waste drum contributes generally to a strong underestimation of the activity and this in particular for radioactive sources emitting low energy gamma-rays independently of their spatial distribution. In this work we present an improved method to quantify the isotope specific activity of spatially concentrated sources (point sources) in nuclear waste drums. The activity is reconstructed using fits of an analytically derived geometric function to count rates recorded during a drum rotation in SGS. The analytical function describes the count rate expectation as a function of the matrix configuration, source location and rotation angle. Based on the fit results the activities of point sources can be reconstructed much more accurate and reliable compared to conventional methods assuming homogeneous matrix and activity distribution.
40 Nuclear Energy and Management of Radioactive Wastes
ANALYSIS OF URANIUM SUPPLY FROM DOMESTIC RESOURCES
G. Zakrzewska-Trznadel1, K. Frąckiewicz1, B. Zielińska1,1. Herdzik-Koniecko1, P. Biełuszka1, A. Miśkiewicz1, K. Szczygłów1, S. Wołkowicz2, R. Strzelecki2, K. Kiegiel1
institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland 2Polish Geological Institute, Rakowiecka Street 4, 00-975 Warsaw, Poland
One of the basic issues of Polish Nuclear Program is to provide necessary analyses of available domestic uranium resources that can be considered as a reasonable reserve of raw material for production of the fuel for Polish reactors. In the period of 1948-1973, in Poland uranium was mined in Sudetes (Kowary, Podgorze, Radoniow, and Kopaliny-Kletno) and Holy Cross Mountains ("Staszic" Mine in Rudki). About 800 tons of uranium was mined on Polish territory and exported to the former Soviet Union. Detailed geological studies carried out to the late 80's of last century have allowed for good recognition of the possibility of occurrence of uranium practically in all lithological structural units of Sudetes. However, due to low content, widespread fragmentation of mineralization and the strong association of uranium with organic matter this occurrence presents no economic significance. According to assessments of Polish Geological Institute the other deposits of uranium are in the Lower Ordovician Dictyonema shale of Podlasie Depression (NE Poland) with concentration of 75-250 ppm and the most prospective uranium mineralization on Polish territory - the Lower and Middle Triassic rocks of the central parts of Peribaltic Syneclise, where concentrations reach even 1,5% U. The ore material originating from these two resources was selected for laboratory studies carried out in the scope of the present project. The project includes also the studies on the possibility of utilisation of secondary resources of uranium like waste material from copper industry and by-products from phosphorous fertilizer industry. The project schedule embraces all stages of uranium separation from the raw material: crushing and grinding, leaching from the ores, purification and concentration of post-leaching solutions involving many processes and finally precipitation and refinement of the product, the yellow-cake - U3O8. The influence of such process parameters like particle size of solid material, leaching mode (acidic or alkaline), concentration of uranium in the solution, temperature and extraction time on efficiency of uranium recovery was tested. The technological part of the project will be completed with technical and economic analysis.
Acknowledgement: The studies are supported by_PO IG 01.01.02-14-094-09-00 research grant: "Analysis of the possibility of uranium supply from domestic resources ".
41 NUTECH-2011
CONCENTRATION OF LIQUID RADIOACTIVE WASTE USING BIOPOLIMER-ENHANCED ULTRAFILTRATION
G. Zakrzewska-Trznadel, A. Miskiewicz, L. Fuks, K. Kulisa
Institute of Nuclear Chemistry and Technology, Warsaw, Poland
Radioactive liquid wastes originating from production and application of radioisotopes contain radionuclides that are predominantly small metal ions like Sr or Co2+. This ions can be removed from radioactive solutions by such membrane methods like reverse osmosis or ultrafiltration if they are large enough to be retained by the membrane. Contrary to reverse osmosis, ultrafiltration is a process that does not need high pressures to be applied. Furthermore, it can involve ceramic or metallic membranes, which are chemically and thermally resistant. They exhibit also high resistivity to ionising radiation that makes them suitable for various applications in nuclear industry. However, when ultrafiltration membranes are applied for retention of metal ions present in low-level radioactive wastes, such small species have to be formerly bound with macromolecular compounds to form complexes that can be easily retained by the membrane. The sorbents like biopolymers in their natural form are inexpensive and abundant materials. They can be produced by different ways: direct extraction from plants (alginates, cellulose) and animal organisms (chitin, chitosan) or they can be synthesized. In the present work biopolymers were tested as potential sorbents for 85Sr and 60Co and as complexing agents for Sr2+ and Co2+ ions in water solutions. Biosorbents based on alginic acid obtained from marine algae, and its derivatives like sodium or calcium alginates were applied in the experiments. The sorbent had solid, granular (calcium alginate) or dispersed, soluble in water (sodium alginate, alginic acid) forms. In the beginning of experiments sorption conditions of Co2+ and Sr2+ ions and radionuclides of 60Co and 85Sr on biosorbents were tested. The influence of pH and ionic strength on the efficiency of sorption was studied. Experiments showed that after 10 minutes of the contact with the biosorbent sorption equilibrium was achieved. The metal complexes formed could be filtered with ultrafiltration membranes, which resulted in high concentration of radioactive substances in retentate. Filtration experiments were carried out with Amicon cell equipped with polysulphone membranes, medium pore size of 10D. Retention of cobalt ions using alginic acid reached 90%, and with sodium alginate ca. 86% at pH 7^8. The retention of strontium ions in the same conditions was 78% and 86%, respectively. Above pH 10 precipitation of hydroxides occurred. The sorption of Co2+ on calcium alginate was 80%, and sorption of Sr2+ slightly exceeded 80%. Increase of salinity of the initial solution resulted in decrease of Co2+ and Sr2+ ions retention. The experiments with 60Co solutions of different salinity and with continuous-mode apparatus were performed. The cross-flow system with different ceramic membranes with 1^-20 kD medium pore size was used. The highest decontamination factors in ultrafiltration/complexation process were obtained when the membrane 1 kD was applied. For 60Co radionuclide at pH=8 decontamination factors were about 10. It was proved that alginate biosorbents can remove radioactive substances from radioactive wastes originating from nuclear industry. In dispersed form they can enhance the ultrafiltration process forming the high-molecular complexes with radionuclides present in the wastes. Acknowledgement: The studies were supported by the National Centre for Research and Development (NCBiR) Research Grant No. R05-058 06/2009. 42 Nuclear Energy and Management of Radioactive Wastes
DESIGN INNOVATIONS IN MANAGING a-CONTAMINATED UNSERVICEABLE GLOVE BOXES
DS Sandhanshive1, SR Shendge1, PP Mazumdar1, A. Kumar2, MNB Pillai2
devendra_barc@yahoo. com
1Technology Development Division, Bhabha Atomic Research Centre, Trombay, Mumbai, India 2Radio-Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai, India
With the maturing of nuclear industry, there is an added burden on the Back End of fuel cycle. Radioactive facilities and radiological laboratories, commissioned decades ago, are in the need for refurbishment or shutting down. This has kept the waste managers, the world over, busier than ever in finding out solutions towards safe handling and disposal of different types and categories of radioactive wastes as an essentiality for environmental remediation. In the Indian context, several Pu-contaminated Glove Boxes were occupying premium storage space in radiological laboratories pending a safe and viable solution for their final management. The decision to dismantle these unserviceable glove boxes had to address the possibility of spread of Pu-contamination in working areas. This appeared to suggest that such glove boxes could have been best dismantled in-situ rather than shifting them to another location. In India's radiological laboratories in Mumbai, such an option was indeed considered. However, the availability of space and headroom in these facilities was too inadequate to effectively launch any dismantling campaign to the complete satisfaction of facility owners, safety regulators and the waste managers. Finally the Glove Boxes were removed from their existing locations and stored in a facility under surveillance. This paper describes the innovative methods employed in managing such unserviceable Glove Boxes and steps taken in that direction. The first step consisted of in-situ encasement of individual Glove Boxes, encountering the challenges of low head room and space congestion in these laboratories with cognizance to regulatory requirement related to radiation safety. The second step was removal, transfer and placement of encased Glove Boxes in a dedicated facility under continuous surveillance. The glove boxes will remain stored in this facility until arrangements are completed for dismantling and volume reduction in another facility which is under design. The final step is the development of an appropriate technique for dismantling/cutting of Glove Boxes in an alpha-tight facility constructed to prevent airborne activity, collection of cut pieces and storage in manageable containers. First two steps in the overall management of glove boxes have already been successfully completed and the third, comprising of the development and design of a dedicated cutting facility is underway. While the design and in-situ handling of Glove Boxes and the engineering efforts of the first two steps have been adequately detailed in this treatise, the contents of the paper are largely devoted to describing the possible options for cutting/dismantling/remote-handling of the Glove Boxes. The description also includes hands on evaluation of tools and gadgets in a full-scale pilot set-up with a view to incorporating the most credible choice in an upcoming active facility.
43 NUTECH-2011
DETERMINATION OF SCALING FACTORS FOR DIFFICULT TO MEASURE NUCLIDES IN SPENT RESINS FROM CERNAVODA NNP
R.Toma1, I. Prisecaru1, C. Dulama2
'University Politehnica Bucharest, Romania 2Institute for Nuclear Research Pitesti, Romania
Radiological characterization represents one of the basic processes in radioactive waste management activities, through which it can be achieved the goal of a high quality final product realization for disposal. In order to properly characterize the radioactive waste packages it is required to know the amounts and concentration of specific radionuclides in the waste package. Many of these specific radionuclides are difficult to measure from outside of the package, as they are alpha and beta emitting radionuclides, and they require laborious and complex radiochemical separation methods, which are not practical for large amounts of waste packages originating from the same waste stream. To avoid performing such radiochemical separation methods, for a specific waste stream, the difficult to measure radionuclides can be correlated to specific radionuclides which are easy to measure from outside of the package [1]. The difficult-to-measure nuclides of primary interest are those with very long half-lives which will persist in a disposal site long after the period of institutional control. Their declaration is often important for the assessment of the health and safety for future uses of the disposal site. The information about the activity concentration and total activity are also required for the transport of radioactive material [2]. This paper aims to develop the theoretical scaling factors for some difficult to measure nuclides which are present in the spent resins from Cernavoda NPP, Romania. This power plant is a CANDU 6 type whith heavy water as coolant and moderator. Moreover, the moderator and the coolant are separated and form two different systems. The spent resins originate from ion exchangers from different purification systems, such as Primary Heat Transport Purification System, Moderator Purification System, Heavy Water Clean-up System. Due to the high concentration of C-14 the spent resins form the Moderator Purification System are stored separately. In Romania, the scaling factor determination methodology has not been established yet and it represents a necessity in order to prepare the radioactive waste packages for final disposal.
'. *** (2009) Determination and use of scaling factors for waste characterization in nuclear power plants, IAEA Nuclear Energy Series No. NW-T-1.18 2. *** (2005) Nuclear Energy - Nuclear Fuel Technology - The Scaling Factor method to determine the radioactivity of low and intermediate level radioactive waste packages generated at nuclear power plant, International Standard, ISO/DIS 21238
44 Nuclear Energy and Management of Radioactive Wastes
IMPLEMENTING PUBLIC PARTICIPATION APPROACHES IN RADIOACTIVE WASTE DISPOSAL
G. Zakrzewska- Trznadel
Institute of Nuclear Chemistry and Technology, ul. Dorodna 16, 03-105 Warsaw
The draft of Polish Nuclear Power Program scheduling the activities necessary for successful implementation of nuclear power in the country was published in 2010. According to the Program the first nuclear power plant will be put in operation around the year 2020. The construction of the NPP implicates the elaboration of radioactive waste management strategy for Poland and first of all, the deployment of new repository for low and medium level radioactive waste, which will be a place for disposal and storage of the waste originating from NPP. The year of 2020 is the time of closure of Rozan site operated since 1961. The new, modern repository housing both institutional waste and waste from NPP will be a necessity and primary task in the plan for radioactive waste management in the country. As one of the few European countries without nuclear power Poland needs an active participation in international initiatives and programmes such as EURATOM research projects. One of such initiatives is the project "Implementing Public Participation Approaches in Radioactive Waste Disposal", which is implemented now under the theme Fission (Research activities in support of implementation of geological disposal), in which two Polish partners, namely Institute of Nuclear Chemistry and Technology and Institute of Atomic Energy take part. The Polish participation is related to the plans of siting of a new repository for low and intermediate level radioactive waste in the country with the perspective of further implementation of geological disposal. Finding appropriate sites for disposal of radioactive waste, especially high level waste and spent fuel, is a very controversial task, not only from technological but also from sociological point of view. To enhance the public participation in decision-making process several approaches like RISCOM model were elaborated and employed in the countries with well- developed nuclear power technologies. The use of advanced models for communicating the society to build the public acceptation and confidence concerning the radioactive waste disposal may be helpful for Poland - the country entering the nuclear energy pathway. The RISCOM model that builds confidence and transparency in the socially sensitive fields, perceived as a potential source of risk, will be implemented in Poland in the scope of EC-7FP IPPA project.
Acknowledgment: The research leading to these results has received funding from the European Atomic Energy Community's Seventh Framework Programme FP7/2007-2011 under Grant Agreement n° 269849
45 NUTECH-2011
METHODS OF IMMOBILIZATION OF RADIOACTIVE ELEMENTS IN SYNROC MATERIALS
T. Smoliński, A. Deptuła, A.G. Chmielewski
Institute of Nuclear Chemistry and Technology, Dorodna 16 str., 03-195 Warsaw, Poland
Nowadays, in the era of expanding nuclear energy, main controversy raises the problem of security of nuclear power plants and storage radioactive waste. Energy from nuclear power plants should be considered safe, without aggravating the environment. To meet this requirement is necessary to find solutions to enable the transformation of dangerous radioactive waste to such a form that they can be safely stored. The solution to this problem may be Synroc type materials which were developed by A.E. Ringwood [1]. Synroc is a kind of particular "synthetic rock" created for the safe storage of radioactive waste. It's an advanced ceramic comprising geochemically stable natural titanate minerals which occur naturally in the earth's crust. These materials allow to incorporate into their crystal structures almost all of the elements present in high-level radioactive waste. Synroc can take many forms which depend on the type and form of waste. In the original form Synroc contains in its structure hollandite (BaAl2Ti6O16) zirconolite (CaZrTi2O7) and perovskite (CaTiO3). Perovskite and zirconolite in their structure immobilize long-lived actinides such as plutonium (Pu), perovskite immobilize mainly strontium (Sr) and barium (Ba). Hollandite is mainly used for incorporation of cesium (Cs), potassium (K), rubidium (Rb) and barium (Ba). Often to the composition of Synroc is added rutile (TiO2) for increased resistance to leaching. The most common method of production of Synroc is synthesis of compounds of titanium in solid-phase. Into obtained matrixes, there are added radioactive waste elements. After that, material is pressed and sintered. An alternative solution synthesis of the various phases Synroc seems to be a sol-gel method. It allows the direct incorporation into the mineral structure of radioactive elements during its formation. Obtained a homogeneous distribution of the components, reduce the sintering temperature and increases resistance to leaching of radioactive elements. In Institute of Nuclear Chemistry and Technology original, patented methods of the sol-gel are used to synthesis Synroc materials with built-in structure surrogates of radioactive elements.
1. RingwoodA.E., (1982) Immobilization of Radioactive Wastes in SYNROC American Scientist; Mar/Apr82, Vol. 70 Issue 2, p201
46 Nuclear Energy and Management of Radioactive Wastes
PHOSPHORANE MINERALS AND PHOSPHORIC ACID AS POTENTIAL SOURCES OF URANIUM
G. Zakrzewska-Trznadel, A. Oszczak, P.Biełuszka
Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland
Raising role of the nuclear power industry, including governmental plans for the construction of the first NPP in Poland, creates increasing demand for the uranium-based nuclear fuels. As a result, one looks for the minerals with the uranium content even lower than 1000 ppm (e.g. phosphorites). Phosphorites, are the sedimentary rocks of the chemical or organic origin. They are formed as the calcium phosphorane precipitates of from the salt water, often followed by the diagenetic processes, or from accumulation of different animal remains (mainly the bones). Among the largest phosphorite deposites are these in the Northern America, Ukraine, Russia, as well as in Morocco, Algeria, Tunisia, Egypt and Mauretania. It was found that optimal conditions for uranium leaching from the phosphorites may be achieved by using one from the following mixtures:
• Alcaline procedure: 0.1 M NaHC03 / 0.1 M Na2C03 containing Mn02 as an oxidant [1], Samples of the Moroccan phosphorites pretreated in 550 °C or in the natural form were
• Acidic procedure: 10 % H2SO4 aqueous solution with MnO2 as an oxidant. Samples of the Moroccan phosphorites pretreated in 550 or in the natural form were leached for 6 h in 30 °C. Uranium content in the samples as well as in the solutions was determined by the ICP MS. For both procedures (In the case of the first / the second procedure) kinetics of the process was studied for differently grained materials. The research included also uranium extraction from phosphoric acid, which is produced in Z. Ch. "POLICE" S.A. from imported phosphate rocks. The liquid-liquid extraction was carried out with two types of phosphoric acid, from Morocco and Tunisia. Before the extraction the 6 acid was oxidized by 15% H2O2 for 1 hour to avoid reduction of U + ions. The volume ratio of two phases (Vw/Vorg) was like 1:1, 2:1 and 3:1. The efficiency of extraction was similar in all cases: after five minutes the extraction of uranium reached almost 100%. As an extracting agent a mixture of two compounds, namely di(2-ethylhexyl)phosphoric acid (D2EHPA) and trioctylphosphine oxide (TOPO) in 0,5 M:0,125M proportion was applied. The kerosene was used as a dissolving agent for the mixture of extractants. After preliminary study, the process will be carried out in the membrane contactor. The membrane extraction will be considered as a one of alternative processes for cleaning and concentration of uranium solutions after leaching from ores and phosphorites. As a continuation of research, the study on selection of optimal conditions for leaching applied to Tunisian and Syrian phosphorite materials are planned. Ammonium carbonate aqueous solutions seem to be promising mixtures for solid-liquid extraction. Acknowledgement: Presented studies were carried out within the POIG Project "Analysis of the possibility of uranium supply from domestic resources" (PO IG 01.01.02-14-094-09-00). 1. J. Steffens (1977) Studies on the Patents concerning Uranium Recovery from the Phosphoric Acid Solutions, D-8032. 2. E.T Romero-Guzman, E. Ordonez- Regil, G.Pacheco-Malagon (1995)Uranium leaching from phosphate rock, J. Radioanal. Nucl. Chem., 201, 313-320.
47 NUTECH-2011
PRELIMINARY PROPOSAL FOR RADIOACTIVE LIQUID WASTE MANAGEMENT IN A BRACHYTHERAPY SOURCES PRODUCTION LABORATORY
Carla D. Souza1, Roberto Vicente1, Maria Elisa C. M. Rostelato1, Carlos A. Zeituni1, Jõao A. Moura1, Eduardo S. Moura1, Fábio R. Mattos1, Anselmo Feher1, Osvaldo L. Da Costa1, Estanislau B. Vianna1, Laércio de Carvalho1, Dib Karan Jr.2
1 Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN - SP Professor Lineu Prestes Avenue, 2242, 05508-000 São Paulo, SP - Brazil. 2 Escola de Artes, Ciências e Humanidades (USP- leste), Rua Arlindo Béttio, 1000 Cep: 03828-000 São Paulo, SP - Brazil.
Malignant tumors are responsible for a high death rate in the entire world population (1). Prostate cancer is the third most common among men, after skin and lung. The treatment using permanent Iodine-125 seed are too costly, preventing the use in large scale (1) (2). A multidisciplinary team was formed to develop a source of Iodine-125 and assemble a national facility for local production. For the production correct implementation, a plan for radiological protection that has the management of radioactive waste fully specified are necessary. This work has developed an initial liquid waste management proposal. The most important Iodine-125 liquid waste is generated in the first phase of the process, radioactive material fixation. The initial proposal is that the waste is deposited in a 20 L container (2 years to fill). The final activity of this container is 4.93 x 1011 Bq. According to the discharge limits presented in the brazilian's regulation CNEN - NE - 605 - Management of radioactive wastes in radioactive facilities (3) this waste could safely be release to the environment in 3.97 years. In the other hand,if a minimization waste policy will be implemented, the production could becomes more efficient and cheaper. Waste storage at 25 L containers and changing some production parameters results in 3 years waste to be eliminated in 3.94 years. This new plan will optimize the materials used and diminished the waste generation facilitating the management, contributing to a cheaper product.
1. MINISTÉRIO DA SA ÚDE. INSTITUTO NACIONAL DE CANCER. Incidência de Cancer no Brasil. Available at:
48 Nuclear Energy and Management of Radioactive Wastes
PROMPT GAMMA CHARACTERIZATION OF ACTINIDES
C. Genreith1, M. Rossbach1, E. Mauerhofer1, T. Belgya2
institute for Energy- and Climate Research - Nuclear Waste Management and Reactor Safety, Forschungszentrum Juelich GmbH, 52425 Juelich, Germany institute of Isotopes HAS, Dept. of Nuclear Research, H-1525 Budapest, POB 77
Due to their pronounced role in nuclear waste management transuranic actinides are of particular interest in research and technology. Chemical and mineralogical behavior has been investigated widely as a homologue element series of the rare earths. However, analytical quantification of minor actinides (Np, Am, Pu, Cm etc.) in mixed waste forms is still hampered and, due to their potentially hazardous nature, not well developed. Radiotoxicity, longevity and decay radiation render these elements interesting for development of analytical tools for their quantification [1]. Unlike other destructive wet-chemical techniques such as atomic spectrometric or mass spectrometric techniques we like to develop a direct, non-destructive approach based on prompt gamma emission after thermal neutron capture. Nuclear analytical techniques (direct a-, ß-, or y-counting) are often not adequate because of long half-life or low gamma ray intensities emitted from the nuclides. High energy gamma rays promptly emitted after neutron absorption may offer an alternative to identify and quantify transuranic actinides in complex matrices. In a first attempt a few selected actinides (Np-237, Pu-242) were irradiated in the well collimated beam of the Budapest Research Reactor with 108 n s-1 cm-2 cold and thermal neutrons to identify the prompt gamma signature of the nuclides, determine differential cross sections, and optimise the counting technique. Experimental results will be compared with theoretical data used in numerical simulation. The identification and quantification of actinides in large samples irradiated with 14 MeV neutrons in a graphite cell will be studied using MCNP5.
1. Wolf SF (2006) Trace Analysis of Actinides in Geological, Environmental, and Biological Matrices. In: Morss LR, Edelstein NM, Fuger J, Kratz JJ (eds) The Chemistry of the Actinides and Transactinide Elements, 3rd ed, Springer, Dordrecht, The Netherlands, Volume 5, Chapter 30, pp. 32 73-3338
49 NUTECH-2011
REDEFINITION OF LARGE LOSS OF COOLANT ACCIDENT (LOCA) IN CONTEXT OF SEISMIC EVENT
K. Demjancuková
[email protected] University of West Bohemia, Faculty of Mechanical Engineering, Department of Power System Engineering, Pilsen, Czech Republic
Recent events that happened in Japan (March 2011) upset the general public and pointed out again to the problem of safety of all present nuclear power plants (NPPs). When solving the problem of nuclear safety of NPP's, various extreme effects have to be involved. Even if these events are highly improbable, they can induce unexpected impacts and vibrations and consecutively environmental, health and biological hazard and naturally a heavy economic loss. The limiting condition for the emergency core cooling system (ECCS) requirements is provided by the double-ended-guillotine break (DEGB) criterion of the largest primary piping system in the nuclear power plant (NPP). The ability of the coolant to remove heat from the fuel is lost - even small losses of fluid may have important consequences. Transition break size (TBS) is a break of area equal to the cross-sectional flow area of the inside diameter of specified piping for a specific reactor. The specified piping for a pressurized-water reactor (PWR) is the largest piping attached to the reactor coolant system, for a boiling-water reactor (BWR) it is the larger of the feedwater line inside containment or the residual heat removal line inside containment. [1] US NRC reached the conclusion that the rupture of primary circuit pipeline with a 850 mm diameter is improbable. Expert elicitation method determined the TBS for PWR within the range of 305 - 356 mm (for the pipelines connected to the main circulation loop) and for BWR of 500 mm, controlled primarily by the surge line and is expected to have a frequency less then 10-5 per year. If the calculation demonstrates the probability less than 10-5 than we will conclude that the influence of large LOCA by a seismic event doesn't exist. The purpose of this paper is to present the main idea and basic assumptions for the assessment of the range of transition break size influenced by a seismic event and also to deal with the problem of seismic data inputs.
1. Tregoning, L. R, Scott, P. M., Abramson, L. R., Chokshi, N.: LOCA Frequency Evaluation Using Expert Elicitation. SMiRT 18, Beijing, China, August 7-12, 2005. 2. U.S.NRC: Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process. Main Report. NUREG-1829, Vol. 1. April 2008. 3. U.S.NRC: Seismic Considerations For the Transition Break Size. NUREG - 1903. February 2008. 4. Knief. R A., Nuclear engineering: Theory and Technology of Commercial Nuclear Power, 2nd ed., American Nuclear Society, 2008, ISBN978-0-89448-458-2.
50 Nuclear Energy and Management of Radioactive Wastes
STUDY OF THORIUM - URANIUM FUEL CYCLE
P. Kalbarczyk, H. Polkowska-Motrenko, E. Chajduk
Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland
Thorium-uranium fuel cycle has recently aroused growing interest because it provides a way of more economic nuclear energy production with lower radioactivity waste comparing to uranium fuel cycle. 232Th is a fertile material. During neutron irradiation, 232Th forms 233Pa - the precursor of the fissile nuclide 233U. The Th-U fuel cycle is also attractive for molten salt reactors of Generation IV type. To explore possibilities of thorium and thorium-uranium fuel 233 233 future use in molten salt reactors, it is necessary to develop methods for of Pa and U as well as uranium isotope ratio determination in irradiated fuel. It is necessary in order to enhance accuracy of nuclear data and models. The need for improvement of nuclear data and models concerning thorium fuel cycle is emphasized in literature [1-3].
The INCT works on analysis of irradiated ThO2 have been carried out. Ten samples of ThO2 of 100 mg mass each were irradiated in nuclear reactor MARIA in thermal neutron flux 14 2 1 233 10 cm s for 250 hours. The samples were cooled for 270 days and Pa has been determined by gamma-ray spectrometry. For determination of 233U and other uranium isotopes inductively coupled mass spectrometry method has been developed. Neutron irradiated ThO2 has been dissolved using THOREX mixture and then 233Pa, U and Th were separated using column extraction chromatography. Uranium isotope ratio has been determined in separated pure uranium fraction by ICP-MS method. Acknowledgement: This work was supported project POIG.Ol.03.01-00-076/08-00, „Analysis of thorium usage effects in an energetic nuclear reactor "
1. International Atomic Energy Agency, Thorium fuel utilization: options and trends. IAEA- TECDOC-1319, Vienna, 2002 2. International Atomic Energy Agency, Thorium fuel utilization: potential benefits and challenges. IAEA-TECDOC-1450, Vienna, 2006 3. D. LeBlanc (2010) Molten salt reactors: a new beginning for an old idea. Nucl. Eng. Des. 240 1644-1656
51 NUTECH-2011
THE DEPLETION ANALYSIS OF HELIOS EXPERIMENT USING MCB CODE
M. Oettingen1'2, E. D'Agata2, Ch. Döderlein2, K. Tućek2, J. Cetnar1
1AGH-University of Science and Technology, Al. Mickiewicza 30, Krakow, Poland 2European Commission, Joint Research Centre, Institute for Energy - P.O. Box 2, 1755 ZG Petten, The Netherlands
The considerable reduction of mass and radiotoxicity of the spent nuclear fuel inventory may be achieved by incinerating Minor Actinides (MAs) in fast spectrum reactors. An irradiation experiment, called HELIOS, was designed in the frame of the EUROTRANS project during the 6th Framework Program of the European Commission. The main aim of the experiment was to gain knowledge about the in-pile behavior of Uranium free fuels containing Americium. The major interest of the experiment was the influence of microstructure and temperature on fuel swelling and He release. Two different approaches were applied to provide He release from the initial stage of irradiation: • creation of open porosity with release paths; • introduction of Pu in order to increase power and, as a result, irradiation temperature. In total, five different fuel samples were irradiated in High Flux Reactor Petten (HFR) at the in-core position G7 from April 2009 to February 2010, what gives about 240 equivalent full power days. The main isotope contributing to radiotoxic inventory is the long-lived Am241. Thereby, all fuel samples consisted of an Am241 dispersed in different inert matrices. The Pu was only present in sample 3 and 5. Sample 1 contained Ceramic-Ceramic (CerCer) type fuel pellets made of the AmOx particles heterogeneously mixed with the MgO inert matrix material. The americium in samples 2 and 3 was incorporated in crystal lattice of the inert matrix forming solid solution with Zr and Y. The fuel samples 4 and 5 were Mo based Ceramic-Metallic (CerMet) composite pellets with a spherical particles diameter of about lOOnm. The focus of our studies is to present an advanced depletion analysis of the HELIOS experiment by means of the Monte Carlo Continuous Energy Burn-up code (MCB). MCB was used mainly to calculate nuclide density evolution in nuclear reactor cores. We present new capability of MCB to investigate the burn-up of nuclear fuel samples irradiated in the HELIOS experiment. In our calculations, we considered the fuel samples inside a spherical neutron source. As an input parameter we use neutron spectrum in XMAS 172 energy group structure and neutron source strength, obtained in previous neutron transport calculations. The transport runs include effects of neutron flux alternation, due to geometrical changes around the experiment, in the nuclear reactor core, whereas fast burn-up runs allow explicit depletion analysis. Moreover, it has been performed a sensitivity analysis to JEF2.2 and JEFF3.1 cross section libraries in terms of the released fission power and the evolution of actinide masses. In addition, we traced the behavior of main fissionable isotopes Am242m and Pu239. The final He production and Am burn-up was also considered.
52 Nuclear Energy and Management of Radioactive Wastes
THE GOLDEN MEAN BETWEEN FUSION AND FISSION - THE FUSION HYBRID
G. Wójcik1 and S.Taczanowski2
[email protected], [email protected]
1PhD student, Faculty of Physics and Applied Computer Science, AGH University of Science and Technology, Cracow, Poland 2Chair of Nuclear Energy, Faculty of Energy and Fuels, AGH University of Science and Technology, Cracow, Poland
Nuclear power plants supply nearly one sixth of the world electric energy production. Though nuclear power is very efficient source of energy it produces dangerous radioactive waste composed of nuclides characterised among others by long decay time and containing also significant quantities of fissile materials. Therefore, spent nuclear fuel must be carefully stored for at least hundreds of years, all this time needing permanent supervision. On the other hand this cumbersome waste contains also important of amounts of energy that should not be simply buried in adequate geological formations. Thus, in spite of the fact that such ultimate disposal of spent nuclear fuel can be recognised as safe - for the energy hungry world trying at the same time to avoid CO2 emissions, this is a hardly acceptable solution. Fortunately, there is another effective approach that is supplied from another akin energy technology, namely - the nuclear fusion. Simultaneously, one has to admit that it till now has not attained yet technological maturity and lengthy and costly investigations still are necessary before nuclear fusion achieves technological development adequate to industrial application. On the other hand the question of nuclear waste needs not a solution right now. A number of decades sufficient for needed progress will have to pass before the problem becomes burning. The idea of some combination of fission and fusion that would join advantages of both these technologies and be used to destroy inconvenient nuclides contained in the spent fuel from fission reactors obviously is not new. Nevertheless, since the hybrid option of fusion power has not enjoyed a support of fusion community the studies performed hitherto have been scarce and not fully satisfactory [1]. Particularly, if the most important requirement the fusion hybrid is facing and that must be fulfilled regards its safety, the accuracy of respective modelling and performed calculations has to be as high as possible. In this view the present study endeavours to meet these demands, while having worked out a heterogeneous model of the hybrid system based upon the mirror concept of fusion reactor. This way the performed evaluations of the system safety with use of such model achieve the needed level of reliability.
1. Taczanowski S. (2007), Premisses for Use of Fusion Systems for Actinide Waste Incineration Int. Conf. EmergingNucl. Energy Syst., ICENES'2007, Istanbul, 06. 2007 www.icenes2007.org/icenes_proceedings/manuscripts.pdf/Session%2010C
53 NUTECH-2011
THE HYBRID SYSTEM FOR LIQUID LOW-LEVEL RADIOACTIVE WASTE TREATMENT WITH APPLICATION OF MEMBRANE PROCESSES
G. Zakrzewska-Trznadel, M. Harasimowicz, A. Miśkiewicz, A. Jaworska
Institute of Nuclear Chemistry and Technology, Dorodna 16, 0-1195 Warsaw, Poland
The aim of the present work is to assess the possibility of treatment of liquid low-level radioactive wastes in the compact installation dedicated to institutions producing small amounts of such wastes. The installation will be based on membrane processes, in some cases enhanced by other methods of treatment that allow 5-10-fold concentration of radioactive species and separation of pure water. The idea of such a multistage, small system comes from rational management of the waste in the place of origin. The aim of the experiments carried out with a small-scale hybrid plant was evaluation of the option of concentration of liquid low- and medium-level radioactive waste using a variety of methods prior to fossilization and final disposal, including: classical filtration, chemical precipitation, ultrafiltration (UF), "seeded" ultrafiltration (SUF), nanofiltration (NF), low- pressure reverse osmosis (LPRO) and high-pressure reverse osmosis (RO). As a final step of waste decontamination the hydrophobic microfiltration membranes were used to perform the process of membrane distillation (MD), or sorption columns for removal the traces of specific radioisotopes that were not eliminated in the previous stages of treatment. The largest volume of liquid radioactive wastes in Poland is produced in nuclear reactor MARIA and in POLATOM at the Institute of Atomic Energy at Świerk/Otwock (about 90 % of total wastes); the rest originates from industry and other users of radioactive materials, for example medical and scientific laboratories. All radioactive wastes collected in Poland are directed to the Radioactive Waste Management Plant in Świerk/Otwock (ZUOP). The main radioactive components of liquid radioactive wastes stored in tanks of ZUOP are such radioisotopes like: 60Co, 85Sr, 106Ru, 125Sb, 134Cs and 137Cs (about 70% of total radioactivity). Specific radioactivity of all ß + y emitters together is 103-105Bq/L and total salt concentration 0.1.2.0 g/L. Radioactive model solutions and real wastes from ZUOP were treated in the single parts of hybrid multistage system described above. The radiochemical purity of the water recovered in this process (more than 95 % of raw waste volume) was kept at the level <10 Bq/L for all ß + y emitters, and <1 Bq/L for a emitters. According to international standards and Polish Atomic Law, this level allows considering this water as non-radioactive waste and discharging it to the city municipal system or reusing for technological purposes.
Acknowledgement: The research was supported by The National Centre for Research and Development (NCBiR) Research Grant No. R05 0058 06/2009.
1. Advances in technologies for treatment of low and intermediate level radioactive wastes. IAEA, Vienna 1994, Technical Reports Series No. 370. 2. Application of membrane technologies for liquid radioactive waste processing. IAEA, Vienna 2004, Technical Reports Series No. 431. 3. Chemical precipitation of radioactive waste, IAEA Technical Reports Series No. 337.
54 Nuclear Energy and Management of Radioactive Wastes
UAE CIVIL NUCLEAR PROGRAMME
K. Szornel
Federal Authority for Nuclear Regulations (FANR), United Arab Emirates
The UAE is pursuing a peaceful, civilian nuclear energy program that upholds the highest standards of safety, security, nonproliferation and operational transparency. The development of a peaceful, civilian nuclear energy program was based on an in-depth evaluation of the UAE's future energy needs. In developing its nuclear energy policy, the UAE government made its peaceful objectives very clear. A policy document released in April 2008 outlined a series of commitments, including the decision to forgo domestic enrichment and reprocessing of nuclear fuel, the two parts of the nuclear fuel cycle that can most readily be used for non-peaceful purposes. Other commitments included operational transparency, non-proliferation, safety and security, partnership with other governments, seeking assistance of expert organizations, and long term sustainability. Consequently the UAE Nuclear Law was signed in October 2009. Under this new law the federal regulatory body FANR was created. The key entities formed to implement the UAE's nuclear energy program are: • Federal Authority for Nuclear Regulation (FANR). An independent federal agency responsible for regulation and licensing of all nuclear energy activities in the UAE with public safety as its primary objective. • Emirates Nuclear Energy Corporation (ENEC). A corporation, wholly Abu Dhabi-owned, charged with developing nuclear power plants within the UAE. ENEC will contract with a primary contractor for the construction of Abu Dhabi's nuclear plants. • International Advisory Board. This advisory body, to include former heads of national regulatory bodies, nuclear industry leaders, and recognized academic authorities, will report directly to the Ministry of Presidential Affairs and provide independent assessments of the status and performance of the various entities associated with the UAE civil nuclear program, as well as analyze progress made in addressing any areas of potential concern. ENEC announced in December 2009 that it had selected a consortium led by Korea Electric Power Corporation (KEPCO) to design, build and help operate civil nuclear power plants for the UAE peaceful nuclear energy program. The KEPCO team includes US-based Westinghouse. The first of the four units is scheduled to begin providing electricity to the grid in 2017, with the three later units being completed by 2020. In December 2010 nine thousand pages application to build first nuclear power station was submitted to FANR and is currently under review.
55 NUTECH-2011
ACUTE TOXICITY ASSESSMENT OF FLUOXETINE HYDROCHLORIDE (PROZAC®) WHEN SUBMITTED TO ELECTRON BEAM IRRADIATION
DRA. Santos1, V.S.G. Garcia, A.C.S. Vilarrubia, S.I. Borrely1
Instituto de Pesquisas Energéticas e Nucleares Av. Lineu Prestes 2242 - Cidade Universitária - Zip code: 05508-000 - São Paulo - SP - Brazil
The large-scale production of medicinal products is directly related to the presence of pharmaceutical drugs in sewage and water. The continuous input of medicines and its residues into the environment especially by sewage and wastewater generates an increasing need of new methods for its treatment and suitable control. The fluoxetine hydrochloride (FH), also known as Prozac®, is an active ingredient used in the treatment of depressive and anxiety disorders [1]. The present study focused on applying the ionizing radiation in order to reduce the acute toxicity of the FH drug solution, under its manipulated formula, to aquatic organisms. Hyalella azteca and Daphnia similis were the organisms used in the biological assays applied for the toxicity studies. It was used a Dynamitron electron beam accelerator and its energy was fixed at 1,4MeV for 5kGy and 10kGy doses [2]. For the calculation of the effective concentration (EC50) it was used the statistic program Trimmed Spearman - Karber. -1 The average values for acute toxicity of FH were 0.59mg.L (EC5 096h) for Hyalella azteca -1 and of 1,44mg.L (EC5 048h) for Daphnia similis. After irradiation of the FH aqueous solution, the following EC50 average values were obtained: 7.81mg.L-1 (5kGy) and 7.97mg.L-1 (10kGy) for Hyalella azteca; 8,46mg.L-1 (5kGy ) and 7.31mg.L-1 (10kGy ) for Daphnia similis. The obtained results revealed the FH as a very toxic compound. These results are confirmed by the EU - Directive 93/67/EEC (Commission of the European Communities) [3]. A significant reduction of the acute effects was obtained when 5kGy and 10kGy were applied.
1. Baldessarini RJ. Drugs and treatment of psychiatric disorders: psychosis and anxiety. In: Hardman JG, Gilman AG, Limbird LE. Ed: Goodman & Gilman's the farmacological basis of therapeutics (1995). New York. McGraw Hill.9;18:399 430. 2. Romanelli, MF, Moraes MCF, Villavicencio ALCH, Borrely SI. (2004). Evaluation of toxicity reduction of sodium dodecyl sulfate submitted to electron beam radiation. Radiation Physics and Chemistry.71:411-413. 3. Blaise C, Gagné F, Eullajfroy P, Férard JF. (2006). Ecotoxicity of Selected Pharmaceuticals of Urban Origin Discharged to the Saint-Lawrence River (Québec, Canada): A Review. Brazilian Journal of Aquatic Science and Technology. 10(2):29-51.
58 Radiation Chemistry
APPLICATION OF RADIATION TREATMENT OF CELLULOSE PULPS FOR PREPARATION OF DERIVATIVES AND MICROCRISTALLINE CELLULOSE
H. Stupińska1, E. Hier2, D. Wawro1, Z. Zimek3, D. Ciechańska1, E. Kopania1
'institute of Biopolimers and Chemical Fibres, Łódź, Poland 2Institute of Atomic Energy POLATOM, Otwock-Swierk, Poland 3Institute of Nuclear Chemistry and Technology, Warsaw, Poland
Institute of Nuclear Chemistry and Technology, Warsaw and Institute of Biopolimers and Chemical Fibers, Łódź realized common research projects which aims were elaborated of radiation methods for modification of properties of cellulose pulps using for derivatives and microcrystalline cellulose manufacturing. The selected cellulose pulps were exposed to an electron beam with energy 10 MeV in linear accelerator. After irradiation pulps underwent the structural and phisico-chemical investigations. The laboratory test for manufacturing carboxymethylocelulose (CMC), cellulose carbamate (CC) and cellulose acetate (CA) with cellulose pulps irradiated dose 10 and 15 kGy have been performed. Irradiation of pulps influenced its depolimerization degree and resulted in the drop of viscosity. However, the expected level of cellulose activation expressed as a rise of the substitution degree or increase of the active substance content in the CMC sodium salt was not observed. In the case of cellulose esters (CC,CA) formation, the action of ionizing radiation on cellulose pulps with dose 10 and 15 kGy enables to obtain of the average values of polimerization degree as required for CC soluble in aqueous sodium hydroxide solution. The properties of derivatives prepared by means of radiation and classic methods were compared. The ecological method of microcrystalline cellulose manufacturing from cellulose pulp employing an effective two-step radiation-enzymatic depolymerization process was elaborated. The properties of cellulose pulp and obtained microcrystalline cellulose were estimated. Polymerization degree, content of crystalline fraction, water retention value, specific volume, whiteness and grain coarseness were determined. The characteristics of the molecular and crystalline structure of microcrystalline cellulose were analyzed by gel chromatography , the microscopic inspection of images and by the wide angle X-ray scattering method. It was documented that the results of preparing microcrystalline cellulose by means of two-step depolymerization depend primarily upon the average cellulose pulp polymerization degree after enzymatic treatment. The microcrystalline cellulose preparation conditions found to be optimal, can be specified as: irradiation of cellulose pulp with a 50 kGy dose followed by enzymatic hydrolysis during 0.5 hour at 50 °C and module enzyme/substrate combination 46 UCMC/g. The depolymerized cellulose pulp is further processed in laboratory scale to form microcrystalline cellulose by milling and drying.. The two-step depolymerization route makes it possible to prepare microcrystalline cellulose with quality indices close to materials used in the pharmaceutical industry, paving the way toward an ecological method of manufacturing microcrystalline cellulose for special uses.
59 NUTECH-2011
CHANGES IN PROPERTIES OF HYDROBIODEGRADABLE FILM BASED ON ALIPHATIC-AROMATIC COPOLYESTERS TREATED BY IONIZING RADIATION
H. Kubera1, K. Melski2, K. Assman1, W. Głuszewski3, Z. Zimek3, N. Czaj a-Jagielska1
1Poznan University of Economics, Faculty of Commodity Science, Department of Industrial Commodity Science, Poland 2Poznan University of Economics, Faculty of Commodity Science, Department of General Chemistry, Poland 3Institute of Nuclear Chemistry and Technology, Warsaw, Poland
Ionizing radiation interacting with animated or inanimate matter may cause multidirectional changes. Investigation and comprehension of the mechanisms and consequences of this interactions allowed to apply the ionizing radiation in many areas, i.e. as a factor modifying the processes of polymerization, in printing ink drying, sterilizing medical implants and utensils, and in food hygiene (Directive 1999/2/EC of 22nd February 1999). Ionizing radiation is characterized by considerable physical hardness through material objects, and therefore, all sterilization processes are carried out in sealed packages, in order to avoid secondary decontamination, and to determine the effectiveness of the process. For many years, there have been studies on the usage of petrochemical polymer materials, used as a radiation-preserved food packaging. These studies resulted in numerous industrial applications. Nowadays, an extensive research is carried out to substitute petrochemical- origin polymers with biopolymers, especially those subject to the process of biodegradation. Optimal research results could be obtained only by the means of close cooperation between material technology and radiation chemistry specialists, as well as by conducting the tests that will support the usefulness of this type of material in contact with food, especially in the context of the consumer health safety. This paper presents the results of the studies on changes of selected quality parameters of hydro-biodegradable aromatic-aliphatic co-polyester (AAC). Radiation treatment was performed at the Institute of Nuclear Chemistry and Technology in Warsaw, using cobalt gamma-ray sources "Issledovatel" (dose rate 0.389 Gy/h) and "GC 5000" (8.5 Gy/h). The material was subjected to the impact of radiation doses of 5, 10, 20 and 40 kGy. Selected parameters of the packaging material, important from the standpoint of a potential application, have been studied: • basic strength parameters in accordance with PN EN ISO 527-3, contact angle measurement using the drop shape method, • global migration PN-EN-1186-1: 2002. The results obtained are presented with the charts of parameter changes as a function of radiation dose. In addition, this paper presents the results of a study on gaseous products of radiolysis, which are the reflection of the degradation processes within the polymer (hydrogen, oxygen, methane and carbon monoxide). These tests have been carried out by the means of gas chromatography, samples exposed to radiation generated by linear electron accelerator, "Elektronika" 10/10 (10 MeV energy and beam power of 10 kW).
Acknowledgements: Project has been carried out with financial support from the funds for science in years 2009- 2012. Research project number NN508393537. 60 Radiation Chemistry
DOSE SENSITIVITY ENHANCEMENT ON POLYMER GEL WITH SUSPENDED GOLD PARTICLES
L.C. Afonso1'2, F. Schöfer2, C. Hoeschen2, L.V.E. Caldas1
Luciana.afonso@helmholtz-muenchen. de
institute of Energy and Nuclear Research, University of Sao Paulo, Sao Paulo, Brazil 2Helmholtz Center Munich, German Research Center for Environmental Health, Neuherberg, Germany
The presence of high-Z materials adjacent to soft tissues, when submitted to irradiation, enhances locally the absorbed dose in these soft tissues. Such effects happen due to the outscattering of photoelectrons from the high-Z materials [1,2,3]. To investigate this effect, polymer gel dosimeters with suspended gold microspheres were used. This study was performed using the polymer gel dosimeter known as MAGIC [4]. The polymer gel was produced and divided into two parts. One part was uniformly mixed with gold microspheres at a concentration of 0.5 % by weight and the other part contained only polymer gel. Each part was poured into test tubes, 4 samples per dose were used, in a range from 0 to 6 Gy. The unirradiated samples were taken as control. The samples were irradiated with X-rays generated at 150 kV, filtered with 4 mm Al and 5 mm Cu, at a constant air kerma rate of 18.9 mGy/min. All samples were read using a Bruker 9.4 T Magnetic Resonance scanner. The transversal relaxation rate R2 (measured in terms of T2=1/R2) from each sample (minus the R2 value from the control) was plotted against the nominal delivered dose to obtain the calibration curves. The samples containing polymer gel with gold microspheres presented approximately 20 % higher change of the R2 value per dose in comparison to the samples containing only polymer gel. This result indicates a dose enhancement factor of approximately 20 %. An analytic estimation predicted a dose enhancement factor of 30 %. The calculations are being refined for a closer description of the experimental parameters.
1. Regulla D, Friedland W, Hieber L et al (2000) Spatially limited effects of dose and LET enhancement near tissue/gold interfaces at diagnostic X ray qualities. Radiation Protection Dosimetry, 90; 1-2:159 163 2. Regulla D, Panzer W, Schmid E et al (2001) Detection of elevated RBE in human lymphocytes exposed to secondary electrons released from X-irradiated metal surfaces. Ratiation Research, 155;744-747 3. Herold DM, Das IJ, Stobbe CC et al (2000) Gold microspheres: a selective technique for producing biologically effective dose enhancement. International Journal of Radiation Biology, 76;10:1357-1364 4. Fong PM, Keil DC, Does MD, Gore JC (2001) Polymer gels for magnetic resonance imaging of radiation dose distributions at normal room atmosphere. Physics in Medicine and Biology, 46;3105-3113
61 NUTECH-2011
EFFECT OF IONIZING IRRADIATION ON TILAPIA (OREOCHROMIS NILOTICUS) SKIN
CAP. Frose1,2, E. Moura1, R.B. Yamaguishi1, E.S.R. Somessari1, C.G. Silveira1, E. Leme1,2, ABC. Geraldo1, J E. Manzoli1
[email protected], [email protected]
instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP, Av. Prof. Lineu Prestes, 2.242, Cidade Universitária, 05508-000, São Paulo, SP, Brazil 2Universidade Paulista -UNIP, Av. Independencia, 412, 18087-101, Sorocaba, SP, Brazil
The culture of tropical tilapia (Oreochomis niloticus), as a non-native freshwater fish specie in Brazil, have had importance in the last decade due its fast growing, fast reproduction characteristics, high tolerance of climate variations in our country and high disease resistance. The high demand of whole fish or fillets is related to its good taste. Althought, its skin, that represents from 4.5 % to14 % of fish weight, is a byproduct that is generally discared or sold at low cost to feed mills. The general animal skin composition comprises protein, water, minerals and fatty matter where the relative portions of these materials depends of upon animal specie, age, breed, feeding and other animal habits. The putrecible raw animal skins can be chemically and physically treated to make it in non-putrecible stabilized material; it results in a soft and flexible polymeric material. The chemical process to obtain this material generally involves a crosslinking of carboxyl groups or amino groups of skin proteins and the chemical reactive specie [1]. Also, physical process as UV irradiation have been successfully employed to crosslink collagenous biomaterials and thus, improved some mechanical characteristics [2]. The goal of this work was to study the tilapia skins exposed to ionizing irradiation from electron beams. The raw skins and the chemically degreased skins were the studied materials. The tensile strenght and elongation at break were the mechanical parameters evaluated. The optical microscopy was used to evaluate some histological characteristics in irradiated and non-irradiated samples. Also, the polymeric product obtained when skins are treated with oxidizing ions were used to compare some results. The tilapia raw skins were kindly available by APTA, a governmental agribusiness technological agency. These skins were scales free, slighted and frosted. The skins were irradiated in atmosphere air on a Job 188 Dynamitron® Electron Beam Accelerator with 1.5 MeV energy under comprised doses of 20 kGy and 40 kGy and dose rates of 2.2 kGy/s and 7.4 kGy/s. The mechanical parameters were measured at a Lloyd LXR tensile tester at a crosshead speed of 10.00 mm/min. Irradiated samples shows high integrity and high tensile strength in comparison to the polymeric product obtained by oxidizing ions reaction. These results are discussed.
1. Knott L, Tarlton JF, Baylei AJ, (1997) Chemistry of collagen crosslinking: biochemical changes in collagen during the partial mineralization of turkey leg tendon. Biochemistry Journal 322:535-542 2. Weadock KS, Miller EJ, Bellincampi LD, Zawadsky JP, Dunn MG, (1995) Physical crosslinking of collagen fibers: comparison of ultraviolet irradiation and dehydrothermal treatment. Journal of Biomedical Materials Research 29;11:1373-1379
62 Radiation Chemistry
EFFECT OF TRANSITION METAL SALTS ON COLLOR, GLOSS AND HARDNESS OF EB-CURED PIGMENTED COATINGS FOR POLYMERIC SUBSTRATES
Marcelo Augusto Gonçalves Bardi1, Mara de Mello Leite Munhoz2, Luci Diva Brocardo Machado2
instituto de Pesquisas Energéticas e Nucleares, Universidade de São Paulo (IPEN/USP) instituto de Pesquisas Energéticas e Nucleares, Comissão Nacional de Energia Nuclear (IPEN/CNEN-SP)
For the last 30 years, UV/EB radiation has been largely used to cure varnishes, inks, adhesives and coatings in order to improve productivity, increase product performance and eliminate hazardous volatile organic compounds (VOCs) [1-2]. Practically, the use of EB curing is more restricted than UV probably because of an apparently higher complexity and investment cost [3]. The desired final material is a cured and cross-linked polymer [4], but, in contrast, thermosetting networks endow excellent thermal and chemical stability, even at environmental conditions after the product is discharged [5]. An alternative can be incorporating compounds (i.e. metal transition salts) that can induce, under controlled conditions, photo-generated holes and electrons that can combine with the surface adsorbed species (e.g., water and oxygen) to form highly reactive radical species such as hydroxyl radicals and superoxide anion [6]. So, the aim of this work is to analyze the effects of the presence of two metallic stearates in EB-cured pigmented coatings formulations by means of changes on gloss, hardness and color as a function of radiation doses in the studied range.
1. Salleh NGN, Yhaya MF, Hassan A, Bakar AA, Mokhtar M (2011) Effect of UV/EB radiation dosages on the properties of nanocomposite coatings. Radiation Physics and Chemistry 80;2:136-141. 2. Bauer F, Decker U, Czihal K, Mehnert R Riedel C, Riemschneider M, Schubert R, Buchmeiser R (2009) UV curing and matting of acrylate nanocomposite coatings by 172 nm excimer irradiation. Progress in Organic Coatings 64;4:474-481. 3. Bénard F, Mailhot B, Mallégol J, Gardette JL (2008) Photoageing of an electron beam cured polyurethane acrylate resin. Polymer Degradation and Stability 93;6:1122-1130. 4. Berejka AJ, Montoney D, ClelandMR, Loiseau L (2010) Radiation curing: coatings and composites. Nukleonika 55;1:97-106. 5. Sangermano M, Tonin M, Yagci Y (2010) Degradable epoxy coatings by photoinitiated cationic copolymerization of bisepoxide with s-caprolactone. European Polymer Journal 46;2:254-259. 6. Sangermano M, Palmero P, Montanaro L (2009) UV-cured polysiloxane epoxy coatings containing titanium dioxide as photosensitive semiconductor. Macromolecular Materials and Engineering 294;5:323-329.
63 NUTECH-2011
ELECTRON BEAM TREATMENT OF EXHAUST GAS WITH HIGH NOX CONCENTRATION A.G. Chmielewski1, A. Pawelec1, J. Licki2
'institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland 2Institute of Atomic Energy POLATOM, 05-400 Otwock-Swierk, Poland
The exhaust gases with high NOx concentration are emitted from diesel engines mounted in the ship. Large marine diesel engines are usually operated on high-sulphur bunker fuel oil which is cheaper. The exhaust gas from such ship contains high concentrations of SO2 and NOx and does not meet the recent IMO regulations and requires the application of a control device for reduction of SO2 and NOx concentrations. The applicability of electron beam treatment technology for purification of such exhaust gases was the main goal of this paper. The experimental study was performed in the laboratory plant with model gas mixture containing high concentrations of NOx in the range from 200 to 1760 ppmv and SO2 in the range from 600 to 1250 ppmv and without gaseous ammonia addition to these gas mixtures. The simultaneous SO2 and NOx removal efficiencies were obtained. The results of parametric tests indicate that SO2 and NOx removal efficiencies from these gas mixtures depend mainly on the following process parameters: absorbed dose, gas temperature at the inlet to process vessel, inlet SO2 and NOx concentrations. Higher inlet SO2 and NOx concentrations yield lower removal efficiencies of both pollutants. The synergistic effect of high SO2 concentration on the NOx removal efficiency was observed in the performed tests.
64 Radiation Chemistry
y-RAY INDUCED VULCANIZATION OF RUBBER COMPOUNDS CONTAINING PRISTINE AND MODIFIED SILICA
D. Dondi1, A. Lostritto2, L. Conzatti3, M. Castellano4, A. Turturro4, S. Bracco5, M. Galimberti2, A. Buttafava1, A. Faucitano1
1 Dipartimento Di Chimica, Université Di Pavia, V. Le Taramelli 12, 27100 Pavia ( Italy) 2 Pirelli Tyres, Milano-Bicocca, Milano ( Italy) 3 Istituto Per Lo Studio Delle Macromolecole - Uos Cnr, Genova ( Italy) 4 Dipartimento Di Chimica E Chimica Industriale, Université Di Genova ( Italy) 5 Dipartimento Di Scienza Dei Materiali, Université Di Milano-Bicocca ( Italy)
The radiation chemistry of rubber/silica compounds has unexploited aspects concerning particularly the role of the direct radiolysis of silica, the nature of its radiolytic intermediates, their interaction with the rubber matrix and their role in promoting the formation of silica-rubber chemical links which are of great importance for the reinforcement mechanism. This work deals with such mechanistic aspects, the kinetics of crosslinking and the mechanical properties of the irradiated SBR/silica compounds. Related to this target are the experiments with silica modified by radiation grafting of polybutadiene olygomers aimed at increasing the free radical reactivity of the silica particle surface and the formation tendency of silica-rubber chemical links. The compounds were prepared by mixing 30-40 % of pristine and modified silica (Zeosil 165 m2/g) with SBR (30% styrene, 45 % vinyl and 25 % aromatic oil); modified silica was prepared (1,2) by impregnation and subsequent radiation grafting of polybutadiene olygomers with average molecular weight Mn = 5000 and 1000 and vinyl double bond content 20% and 45% respectively.y irradiations were performed under nitrogen atmosphere in a 60- Co source with total doses ranging from 100 to 400 kGy. Information concerning the mechanism of crosslinking and silica-rubber chemical links were obtained by EPR spectroscopy and 13-C, 29-Si CP/MAS NMR spectroscopy. The mechanical behaviour of the compounds was investigated by tensile and dynamic-mechanical measurements. In silica a powerful, low activation energy, free valence migration mechanism exists leading to radicals diffusion toward the surface and decay. The species attaining the surface react with absorbed SBR vinyl double bonds thus generating grafted radicals and silica-rubber chemical links. When modified silica is used, the formation of silica-rubber bonds is enhanced by the free radical scavenging properties of the polybutadiene coating. With SBR blends containing pristine silica, the crosslinks yields, as determined by the Flory-Rhener method, grow almost linearly with the radiation dose attaining a value > 2.5x10-4 moles/kg at about 400 kGy (both physical and chemical links). When using modified silica, lower initial yields are observed followed by an acceleration period which ultimately leads to a full recovery of the overall crosslinks yields with respect to pristine silica. The initial lower yields are thought to arise from the attenuation of the silica surface energy by the olygomer grafting. The subsequent acceleration is related to the taking over of the free radical scavenging mechanism leading to silica-rubber chemical links. Mechanical measurements are consistent with this interpretation showing that compounds containing modified silica show lower initial mechanical strength followed by a faster response to the radiation dose. A TEM characterization of the vulcanizates is presented.
1. Dondi D, Palamini C, Pepori F, Buttafava A, Galinetto P, Faucitano A (2009) Nukleonica, 54, 71- 75 2. Dondi D, Palamini C, Buttafava A, Faucitano A, Galinetto P, Nahmias M,.Giannini L, Lostritto A (2010), Macromol. Symp. 296, 38-43
65 NUTECH-2011
HIGH-TEMPERATURE OXIDATION RESISTANCE OF STAINLESS STEEL DOPED WITH YTTRIUM USING ION IMPLANTATION
M. Barla^JTPiekoszewskTI, Z. Werner1, B. Sartowska2, L. Waliś2, W. Starosta2, J. Kierzek2, K. Bocheńska1, R. Heller", R. Wilhelm3, A. Kolitsch3, C. Pochrybniak1'4, E. Kowalska5
Andrzej Sołtan Institute for Nuclear Studies, Świerk/Otwock, Poland 2Institute of Nuclear Chemistry and Technology, Warsaw, Poland 3Helmholtz-Zentrum Dresden-Rossendorf, Dresden, Germany institute of Atomic Energy, 05-400 Otwock/Świerk, Poland 5Military University of Technology, Warsaw, Poland
The addition of some amount of oxygen reactive elements like Y and rare earth elements (REE) Ce, La, Er and others into stainless steels or iron chromium alloys improves their oxidation resistance at high temperature. There are numerous methods of incorporation of REE into steel by surface treatment, e.g.: ion implantation, metal organic, chemical vapour deposition, sol-gel coating, pack cementation, screen-printing, molten-salt electrodeposition. In the present work we intend to use yttrium as an active element which will be incorporated into 304, 316 and 430 stainless using conventional implantation with MEVVA type of Y ion source. The samples will be examined by Scanning Electron Microscopy (SEM), Energy Dispersive X-ray Spectroscopy (EDX), X-ray diffraction (XRD) and crucially important Rutherford Back Scattering (RBS) measurements and subjected to oxidation in air at a temperature of 1000°C for a period 100 h. Results obtained with the use of HIPPB method and ion implantation will be compared and discussed.
66 Radiation Chemistry
HUMIDITY EFFECT ON RADIATION CHEMICAL DEGRADATION YIELD OF CHITOSAN
S. Benamer1, M. Mahlous1, D. Tahtat1, A. Nacer Khodja1, Murat §en2
1 Division of Nuclear Applications, Centre de Recherche Nucléaire d'Alger, BP-399 Alger-Gare, Algeria 2Hacettepe University, Department of Chemistry, 06532, Beytepe, Ankara, Turkey
This study focuses on the humidity effect on radiation degradation of chitosan in solid state. Chitosan samples having different humidity rates were irradiated at different doses with gamma rays in air at ambient temperature. The changes in molecular weight of chitosan were followed by viscosity methods and size exclusion chromatography (SEC). Radiation chemical degradation yield G(s) was used to evaluate gamma radiation scission of chitosan. It was found that the humidity rate was an important factor controlling the G(s) and degradation rate. the reduction in humidity rate leads to a significant increase in chitosan molecular weight, however the radiation chemical yield increases from 0.81 for dried chitosan (0% humidity) to 1.44 for samples having 100% of humidity rate.
67 NUTECH-2011
INFLUENCE OF IONISING AND UV RADIATION ON TEMPLATE DEPOSITED MICROSTRUCTURES OF SILVER HALOIDS
M. Buczkowski, B. Sartowska, W. Starosta
Institute of Nuclear Chemistry and Technology, Laboratory of Material Research Dorodna 16, 03-195 Warsaw, Poland
Nano-/microstructures of different materials synthetized into membrane templates are objects of growing interest because of their unique properties (one-dimentional structures) and connected with it potential application in many fields.[ 1,2,3] In the Institute of Nuclear Chemistry and Technology investigations concerning nano- /microstructures of silver haloids inside track-etched membrane templates have been undertaken. Such membranes are characterized by precisely determined cylindrical pores with typical pore size in the range: (0.2 - 2.5) [j,m. Suitable procedure using chemical exchange reaction for getting above structures have been worked out. Microrods into membrane pores and micrograins on membrane surfaces were deposited. Scanning electron microscopy (SEM) with X-ray microsonde was used for getting morphology as well as stoichiometry (AgCl, AgBr) of obtained samples.[4] It is observed that the colour of the samples practically do not change in case of scattering visible light. This fact is in contrast with darkening of AgCl or AgBr samples prepared in bulk (without using of membrane templates). -60 source) with doses up to about 100 kGy and UV radiation (254 or 366 nm) on the samples mentioned above with deposited silver haloids microstructures was registered.[5] The samples darked after exposition with above radiation and the level of colour changing can be registered by a photometric method. In case of EB or y radiation, forming of nanograins of silver haloids inside membrane pores were observed. It seems that the obtained structures can be used as indicators of UV or ionising radiation.
1. Wade TL, Wegrowe JE (2005) Template synthesis of nanomaterials. Eur. Phys. J.-Appl. Phys. 29:2-22 2. Potiyaraj P, Kumlangdudsana P, Dubas ST (2006) Synthesis of silver chloride nanocrystals on silk fibers. Mat. Letters 61, Issues 11-12:1264-2467 3. Sambhy V, MacBride M, Petersom BR, Sen A (2006) Silver bromide nanoparticle/polymer composites: Dual action tunable antimicrobial materials. J. Am. Chem. Soc. 128:9798- 9808 4. Buczkowski M, Sartowska B, Starosta W (2010) Template synthesis of nano- /microstructures of silver haloids and investigation of their selected properties. Proc. of EMRS Fall Meeting (Section - Nanotechnologia PL) 13-17 Sept. 2010, Warsaw, Poland 5. Chmielewski AG, Michalik J, Buczkowski M, Chmielewska DK (2005) Ionising radiation in nanotechnology. Nucl. Instr. andMeth.in Phys. Res. B236:329-332
68 Radiation Chemistry
KINETICS AND MECHANISMS OF RADIATION-INDUCED DEGRADATION OF HEXACHLOROCYCLOHEXANE IN WATER
Hasan M. Khan, Sanaullah Khan
Radiation Chemistry Laboratory, National Centre of Excellence in Physical Chemistry, University of Peshawar, Peshawar 25120, Pakistan.
Hexachlorocyclohexane (y-HCH), commonly known as lindane, is an organochlorine insecticide which is characterized by high toxicity, persistency, bioaccumulative and long- range transportable nature. Several conventional methods, such as aeration, flocculation, filtration and biodegradation etc. are used for removal of lindane and other POPs. In recent years, a new technology, called advanced oxidation and reduction technology (AORT), which involves the oxidation or reduction of contaminants by reactive free radicals, is emerging as a promising technique for treatment of several organic and inorganic pollutants in water. Mechanism of lindane degradation by advanced oxidation technology using gamma irradiation was investigated in this work. Aqueous solutions of lindane were irradiated on a laboratory scale over a range of 0.15 to 3.7 kGy using a 60Co radiation source. Gas chromatography using Electron Capture Detector (ECD) was used for analysis of lindane in water. Solid phase micro extraction (SPME) fiber along with CTC autosampler was applied for extraction of lindane from water. The lindane degradation efficiency was very much enhanced when the solution was degassed (removing oxygen from the matrix), indicating - aqueous electron (e aq) radicals as the possible reacting specie. The lindane degradation efficiency was, however, reduced in the presence of N2O, a good electron scavenger, which - further supports the role of aqueous electron (e aq) radicals in the degradation mechanism. Effects of other radical scavengers on lindane degradation, such as CO32", HCO3-, NO3", NO2", H2O2, iso-propanol, and ter-butylalcohl were further investigated. The lindane degradation efficiency was enhanced in the presence of CO32- and HCO3- radicals, while the efficiency was decreased in the presence of NO3- and NO2- radicals. The presence of iso-propanol and ter-butylalcohl did not significantly affect the degradation process. The process follows a first order reaction kinetics and the reaction was mainly controlled by the reaction of lindane with - aqueous electron (e aq) radicals. Possible reaction mechanism and reaction intermediates will be discussed.
69 NUTECH-2011
LIFETIME PREDICTION OF CABLES INSTALLED IN NUCLEAR POWER PLANTS BASED ON ANTIOXIDANT DECOMPOSITION IN INSULATIONS
J. Boguski, G. Przybytniak, K. Mirkowski, A. Bojanowska-Czajka
Institute of Nuclear Chemistry and Technology, Dorodna 16 Str., 03-195 Warsaw, Poland
The polymers used for the insulation and jacket materials in electric cables are susceptible to aging and degradation processes induced by many of the stressors. In nuclear power plant service cables might be exposure to low dose rate irradiation, elevated temperatures, humidity etc., that could significantly shorten their service life which in turn can lower the reliability of electrical power systems. The integrity of electric cables is supposed to be monitored through periodic functional inservice testing. Therefore, a lot of research effort and activities are directed towards a better understanding degradation phenomena, and establishing the accurate methods for insulation diagnosis. In order to improve the physical properties and extend the lifetime of cables their insulations contain additives hindering degradation, e.g. antioxidants, that gradually decompose during aging. Their consumption with time might cause unexpected early failures of the electric cables, what disrupts operation of electrical equipment, signal transmission and brings about considerable risk of nuclear accident. In the reported studies the level of H-donors phenolic antioxidants, namely IRGANOX 1035, in model systems was investigated. Polyethylene (MALEN E FGNX, 23-D022) free from any additives was doped with various concentrations of antioxidants and subsequently subjected to accelerated radiation aging in gamma chamber. After optimizing method of antioxidant extraction with methylene chloride, the level of detectable additives was determined before and after irradiation applying HPLC with UV detection. The results were compared with oxidation induction time (OIT) conducted at several selected temperatures from the range of 200 - 230 oC. Both techniques might be used for the estimation of relationship between absorption doses and the level of antioxidants still present in the material and their decrease in the course of the irradiation. A correlation between data obtained by liquid chromatography measurements and OIT was established. Both methods might be utilized to predict the service lifetime of some cables installed in the reactor containment of nuclear power plants.
70 Radiation Chemistry
MECHANICAL EVALUATION OF PVC FILMS MODIFIED BY ELECTRON BEAM IRRADIATION
J.R. Cardoso, E. Moura, E.S.R. Somessari, C.G. Silveira, H.A. Paes, C.A. Souza, J.E. Manzoli, A B C. Geraldo
[email protected], [email protected]
Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP, Av. Prof. Lineu Prestes, 2.242, Cidade Universitária, 05508-000, São Paulo, SP, Brazil
The polyvinyl chloride (PVC) is a technological and low cost polymer. Although this polymer is weather resistant, it presents high sensitivity to high energy irradiation because of the weakness of carbon-chloride bond face to carbon-carbon and carbon-hydrogen bonds [1]. Upon exposure to high energy irradiation, some carbon-chloride bonds are broken to give rise radicals like chlorine and organic chloride that are initiators of two concurrent processes: degradation and crosslinking. The degradation process on PVC is macroscopically observed by discoloration effects, where the material tends to darken or to turn yellow, mainly at a typical sterilization dose of 25 kGy [2]; this process yields hydrogen chloride as a byproduct, it is autocatalytic and it continues after irradiation exposure. By other way, the crosslinking leads to an improvement in thermal resistance and mechanical properties. The aim of this work is to evaluate the mechanical properties of PVC irradiated by electron beam to verify the degradation process. Also, mechanical properties are investigated on styrene grafted PVC by electron beam irradiation using mutual and pre-irradiation methods to verify the mechanical resistance changes of obtained product if grafting process is applied from non-irradiated and from pre-irradiated substrates. The PVC commercial film samples with 210 jam thickness were cut into a dumbbell-shape (gauge length: 25 mm, width: 4.10 mm, area: 0.86 mm2). The grafting media was styrene /butanol-1 mixture in several monomer concentrations (from 10% to 100%). The samples were irradiated on a Job 188 Dynamitron® Electron Beam Accelerator with 1.5 MeV energy. All irradiation procedures were performed in atmosphere air and the irradiation conditions comprised doses from 10 kGy to 100 kGy and dose rates of 2.2 kGy/s and 22.4 kGy/s. The styrene grafted samples were analyzed by gravimetry to determinate the grafting yield; the final values have been averaged from a series of three measurements. The Mid-ATR-FTIR was the spectrophotometer technique used for qualitative/semi-quantitative analysis of grafted samples. The Young's module and tensile strength of pre-irradiated and grafted PVC samples at both methods were measured at a Lloyd LXR tensile tester at a crosshead speed of 10.00 mm/min. e of absorbed dose at pre-irradiated PVC samples, that it indicates degradation process. These mechanical parameters results are discussed to styrene grafted PVC samples.
1. Motysia BS, Zimek Z, Bojarski J, Przybytniak G, Sadlo J (1998) Radiation effects on PVC and PVC compositions. Radiation Chemistry and Physics, 49;54-56 2. Wang S, Zhang Y, Zhang Y, Zhang C, Li E (2004) Crosslinking of polyvinyl chloride by electron beam irradiation in the presence of ethylene-vinyl acetate coplolymer. Journal of Applied Polymer Science 91;1571-1575
71 NUTECH-2011
PRELIMINARY GAS CHROMATOGRAPHY/MASS SPECTROMETRY EVALUATION OF POLYCHLORINATED BIPHENYLS REMOVAL FROM WASTEWATERS BY GAMMA IRRADIATION
M. Virgolici1, I. Dobrica2, I.R. Stanculescu1'2, A.V. Medvedovici3, MM. Manea1, M. Alexandru1, C D. Negut1, M. Cutrubinis1, I.V. Moise1 and C.C. Ponta1
1"Horia Hulubei" National Institute of Physics and Nuclear Engineering (IFIN-HH), IRASM Centre of Technological Irradiations, 077125, Mãgurele (Ilfov), Romania 2University of Bucharest, Faculty of Chemistry, Department of Physical Chemistry, 030018, Bucharest, Romania 3University of Bucharest, Faculty of Chemistry, Department of Analytical Chemistry, Bucharest, Romania
Chronic shortages of water in arid and semi-arid regions of the world and environmental policy regulations have stimulated the use of appropriate technologies in treating wastewater for reuse, for example, in urban irrigation, industrial uses (cooling, boilers, and laundry), gardens and parks, cleaning purposes, etc. Additionally, wastewater treatment is becoming ever more critical in large industrial centres due to diminishing water resources, increasing water disposal costs, and more restrictive discharge regulations that have lowered permissible contaminant levels in waste streams. Industrial effluents often carry chemical contaminants such as organics, petrochemicals, pesticides, dyes and heavy metal ions. The standard biological treatment processes commonly used for wastewater treatment are not capable of treating many of the complex organic chemicals that are found in varying quantities in the wastewaters (e.g. persistent organic pollutants, POPs). Radiation-initiated degradation of organics helps to transform various pollutants into less harmful substances or reduced to the levels below the permissible concentrations. Present work on gamma radiation degradation of polychlorinated biphenyls (PCBs) has been done in the frame of an ongoing IAEA research grant entitled "Extensive Use of Gas Chromatography - Mass Spectrometry for the Characterisation of the Effects of Radiation Treatment on Wastewater" (Contract No. 16426/RO), part of Co-ordinated Project: "Radiation Treatment of Wastewater for Reuse with Particular Focus on Wastewaters Containing Organic Pollutants". To identify most convenient irradiation conditions for wastewater treatment, samples were irradiated at different doses and two dose rates: 2 kGy/h and 30 kGy/h. The dependence of PCBs concentration of dose and dose rate was monitored with gas chromatography coupled with an electron capture detector (GC-ECD). Same samples were screened with gas chromatography coupled with mass spectrometry (GC/MS), in order to identify radiolysis products which may still be classified as persistent organic pollutants. Compound identification was performed via mass spectra deconvolution and retention index with AMDIS software and spectral matching with NIST 2005 GC/MS library. The results of the GC/MS screening will designate the target compounds for routine quantitative monitoring by GC-ECD. A molecular modelling study was developed to predict the most probable radiolysis products and it was correlated with the GC/MS results to speculate some preferential radiolysis pathways.
72 Radiation Chemistry
R&D LABORATORY WITH LINEAR ACCELERATOR FOR RADIATION PROCESSING
M. Fiilöp1, L. Harmatha2, M. Żiśka2, M. Nemec2,1. Benkovsky3
:R&D Laboratory with linear accelerator, Slovak Medical University, Limbova 12, 833 03 Bratislava 2Department of Microelectronics, Faculty of Electrical Engineering and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava, Slovakia department of Pharmaceutical Analysis and Nuclear Pharmacy, F&culty of Pharmacy, Comenius University in Bratislava, Odbojárov 10, 832 32, Slovakia
The R&D laboratory with electron accelerator of Slovak Medical University was established in 2011 aiming at successful radiation technology implementation in some specific areas like radiation sterilization, polymer and semiconductors modification. The nominal technical specifications of the laboratory are as follow: - electron accelerator UELR 5-1C with X-ray converter, 5 MeV electron energy, beam current from 3 up to 200 mA, beam power up to 1 kW, the length of the scanned beam up to 50 cm, - conveyor system with multi passes transfer and turnaround of the product, - reference dosimetry system based on GEX B3 WINdose radiochromic film dosimeters, polystyrene calorimeters and alanine-EPR spectrometry, - process qualification and control in electron beams with the Monte Code MCNPX Staff of the R&D laboratory was prepared several years mainly in framework of IAEA projects : Quality Control Methods and procedures for Radiation Technology (RER/8/010), Standard Feasibility Study for Electron Beam Flue Gas Treatment Technology (RER/8/011) and Enhancing Quality Control Methods and Procedures for Radiation Technology (RER/8/017). In the future the R&D laboratory would like to participate in Technical Cooperation projects of IAEA in the field of advanced electron beam and electric fields generated plasmas technologies for pollutant emission control, harmonization of standardized quality control procedures in radiation technologies and in the proposed Technical Cooperation project "Innovative nuclear technology in medicine - education and training". One of the introductory studies being elaborated at the R&D laboratory is processing of semiconductive material by electron beam. The new generation of semiconductor structures and devices utilizes advanced materials and fabrication procedures. The knowledge of the rise of radiation defects in semiconductive materials is particularly important from the point of view of device reliability. The radiation resistance of structures based on wide band gap semiconductors has been examined on samples irradiated by high energy electrons. Creation and amounts of radiation defects have been observed by means of current and capacitance measurements on Schottky and MOS structures. The high density of radiation defects after irradiation by 5 MeV electrons and doses up to 300kGy was not destructive but it increased the threshold voltage in MOS structures and the leakage current in Schottky structures markedly.
73 NUTECH-2011
RADIATION GRAFTING OF ACRYLIC ACID ONTO CHITOSAN BEADS FOR METAL IONSORPTION
S. Benamer1, M. Mahlous1, D. Tahtat1, A. Nacer-Khodja1, M. Arabi1, H. Lounici2, N.Mameri2
division of Nuclear Applications, Centre de Recherche Nucléaire d'Alger, BP-399 Alger-Gare, Algeria 2 Ecole Nationale Polytechnique d'El-Harrach Alger, Algeria
Radiation-induced grafting of acrylic acid onto chitosanbeads was performed at room temperature. The grafting was confirmed by grafting yield, FTIR and methylene blue coloration. The effect of absorbed dose on grafting was investigated. The removal of heavy metals from aqueous solutions was investigated using natural and grafted sorbent chitosan beads, which have been grafted with acrylic acid by gamma irradiation technique at the dose of 4 kGy and dose rate of 20.64 Gy/min. The grafting yield was over 80% and increase by increasing radiation doses. The characterization of the sorbents was achieved by, particle size, FTIR spectroscopy and swelling at different pH). The ability of modified (Chit-AAc) and unmodified (Chit) chitosan beads as sorbent for Pb and Cd ions in aqueous solution was studied. The sorption behavior of materials was examinated through pH, kinetic and equilibrium experiments. Grafted chitosan beads presented higher sorption capacity for both metal ions than unmodified chitosan beads.
74 Radiation Chemistry
RADIATION MODIFICATION OF ELASTOMERS
W. Głuszewski1, Z.P. Zagórski1, M. Rajkiewicz2
1 Institute of Nuclear Chemistry and Technology, Warsaw, Poland 2Institute for Engineering of Polymer Materials and Dyes - Piastow, Poland.
Radiation techniques allow a convenient way to induce reactive intermediates in materials which, in the case of polymers can initiate formation of cross-links. In this way it is possible to modify favorably properties of many polymeric materials [1]. A classic example of the application of ionizing radiation in plastics processing is the radiation crosslinking of polyethylene, which is irradiated on industrial scale in the production of materials such as shape remembering shrink materials or hot water pipes. The question arises whether similar benefits can be obtained from the radiation modification of elastomers [2]? Uniqueness of application of radiation lies in the fact that the process of modification can be realized at any temperature; usually these are the ambient conditions. Even at ambient temperatures the effects of ionizing radiation can initiate reactions similar to high-temperature chemical processes. This totally different quality of energy is expensive. The advantage of radiation treatment is a simple way to control the amount of the radiation dose of absorbed energy. The communication presents the first results of the synergy effects in crosslinking of elastomers using an electron beam, gamma radiation and classical methods of curing. Original methods of investigation were used to test effects of elastomers radiolysis, i.e. diffuse reflection spectrophotometry (DRS) [3] and gas chromatography [4]. For the first time, DRS absorption spectra of products formed at the surface of elastomers were observed, formed as a result of oxidation induced by ionizing radiation. Using the GC, yield of radiation separated hydrogen was determined, which can be regarded as a measure of efficiency of the crosslinking process. Entanglements induced by radiation in some elastomers, like HNBR are not connected with the release of hydrogen, therefore radiation yield of crosslinking can be twice higher than the radiation yield of hydrogen. Determination of oxygen absorbed by the elastomers was also possible by GC. The results proved to be useful to describe the mechanisms of cross-linking, oxidation and degradation of polymers. Irradiation of samples was done in commercial source of cobalt 60 gamma rays GC 5000 (9 kGy/h) and electron linac - "Elektronika" 10/10 (10 MeV, 9 kW). Spectra were recorded on Perkin Elmer Lambda 7 spectrophotometer with an integrating sphere. Gas chromatograph type GC 2014 by Shimadzu, equipped with thermal conductivity detector, column packed with molecular sieves 5A was applied for the analyses of the gaseous products: hydrogen, oxygen, methane and carbon monoxide.
Acknowledgments: Work done under the research project of the Ministry Science and Higher Education Nr N N209 083838
1. Bik J, Głuszewski W, Rzymski W.M, Zagórski Z.P, (2003) EB radiation crosslinking of elastomers, Radiation Physics and Chemistry 67, 421-423 2. Głuszewski W, Zagórski P. Z, (2010) Procesy radiacyjnego sieciowania polimerów, Tworzywa Sztuczne i Chemia 2, 58-60 (in Polish) 3. Zagórski, ZP, (2003) Diffuse reflection spectrophotometry (DRS) for recognition of products of radiolysis of polymers, Int.J.Polymeric Materials 52, 323-333 4. Głuszewski W, Zagórski ZP, Rajkiewicz M, Mikołajska A, (2010) Od Marii Skłodowskiej-Curie -21 (in Polish)
75 NUTECH-2011
RADIATION MODIFICATION OF THE PHYSICOCHEMICAL AND FUNCTIONAL PROPERTIES OF THE POLYSACCHARIDE FILMS
K. Cieśla
kciesla@orange. ichtj. ~waw.pl
Institute of Nuclear Chemistry and Technology, Dorodna 16 str., 03-195 Warszawa, Poland
Last years the interest increases in substitution of the artificial plastics by natural polymers derived from renewable resources, and for substitution of the methods of production that involve strong chemicals by more friendly for the environment methods. Application of polysaccharides extracted from plant or animal tissues and food proteins, alone or composed with the other biopolymer or with artificial polymer, are the explored possibilities. Using of natural polymers, common in the environment and often constituting the agricultural bi-products, might have furthermore the positive impact on the economy. In particular, the trials are continued of utilization of such compositions for production of biodegradable and edible packaging: films and coatings. Functional properties of biopolymer films are generally inferior as compared to the properties of artificial films. Their possible application required thus appropriate improvement of the properties. Next to modification of the films composition various physical or chemical treatments are tested in purpose of optimization of the films quality. Polysaccharide films generally reveal relatively high tensile strength as compared to these of to the protein films, and sometimes even similar to those of the artificial plastic. However, application of these films is strictly limited, in particular in free standing packaging, because of their brittleness and hydrophilic properties of the constructive polysaccharide. Application of ionizing radiation seems to be the alternative friendly for the environment method as compared to chemical methods. It's potential for improvement of the films prepared basing food proteins as a major component is already relatively well recognized [1,2]. Addition of small amount of selected polysaccharides into the protein system (prior to or after irradiation) was also described. Last years several trials were also carried out for elaboration of the methods based on ionizing radiation for improvement of the packaging material containing polysaccharide as a major component, as an equivalent component in polysaccharide-protein system or composed with the artificial polymer. The trials are carried out for utilization of the agricultural products and bi-products containing polysaccharides. The presentation will constitute a short review concerning application of ionizing radiation for improvement of the functional, physicochemical and structural properties of the potential packaging materials containing polysaccharides in their composition. Short presentation of the author's own results concerning starch based films will be also included [2-3].
1. Cieśla K (2005) Application of gamma irradiation for modification of the properties of biodegradable packaging materials based on protein systems (pol), in: Radiation Technology in Industry, Medicine, Agriculture and Environmental Protection, ed. Department of Physics and Applied Mathematics, University of Mining and Me talurgy (AGH), Kraków 501-506 2. Cieśla K (2009) Transformation of supramolecular structure initialized in natural polymers by gamma irradiation. Institute of Nuclear Chemistry and Technology, Warszawa 3. Cieśla KA, Nowicki A, Buczkowski MJ (2010) Preliminary studies of the influence of starch irradiation on physicochemical properties of films prepared using starch and starch-surfactant systems. Nukleonika 55;2:233-242
76 Radiation Chemistry
RADIATION SYNTHESIS OF PVA/CHITOSAN HYDROGEL FOR WOUND HEALING ENHANCEMENT
A. Nacer Khodja1, M. Mahlous1, D. Tahtat1, S. Benamer1, S. Larbiyoucef, H. Chader2, L. Mouhoub, M. Sedgelmaci, N. Ammi2, M B. Mansouri2, S.Mameri3
1 Division of Nuclear Applications, Centre de Recherche Nucléaire d'Alger, BP-399 Alger-Gare, Algeria laboratoire de Controle des Produits Pharmaceutique, Dely Brahim, Alger, Algeria. 3C.H.U. Beni Messous, Alger, Algeria
Hydrogels based on poly (vinyl alcohol) and chitosan synthesized by gamma irradiation were evaluated as wound dressing material, compared to paraffin gauze dressing and natural healing (cotton gauze) in a burn rat model. Histological analysis, Primary irritation index and Ocular Irritation Indexwere investigated. Wounds treated with hydrogel PVA/chitosan healed at 09th day, but with paraffin gauze dressing and cotton gauze healed up to 16th day. Histological analysis showed that new granulation tissue and epithelialization progressed fastly in wound treated with hydrogel PVA/chitosan. The Primary irritation index of the hydrogel PVA-Chitosan was calculated to be 0.5 and the maximum Ocular Irritation Index (O.I.I) was calculated to be zero, these indicate that the hydrogel PVA/chitosan can be considered not irritant to the skin.
77 NUTECH-2011
RADIATION SYNTHESIS OF SILVER MICRO- AND NANOPARTICLES EMBEDDED IN COTTON FABRIC
D.Chmielewska, W. Starosta
Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland
There have been a lot of developments using nanotechnology in the textile industry recently. Different kinds of antimicrobial finishing have been applied to protect garments against harmful microorganisms [1,2]. Among various antimicrobial agents silver has attracted a great deal of scientists' attention because of its complex mechanism of interaction with cells that results in very wide spectrum of antimicrobial activity against bacteria and fungi and low toxicity to mammals' cells simultaneously. In this work an effective antimicrobial agent for cotton fabric was introduced. The silver loaded cotton composites was prepared through immersion of fabric in silver salt solution and following electron beam irradiation. Obtained cotton fabrics with silver indicated an antimicrobial activity against both Staphylococcus aureus and Escherichia coli (fig.1) bacteria as well as Aspergillus niger and Rhodotorula mucilaginosa fungi. In this work we study influence of irradiation doses and silver salt initial concentrations on silver formation and agglomeration within cotton matrix. The mechanism of the process is based on ion adsorption - radiation induced metal ion reduction - diffusion forced aggregation phenomena, so the methods of nanostructures stabilization are discussed. SEM-BSE, EDS and XRD methods were applied to determine silver particle sizes, distribution and concentration in the samples. Influence of silver particles concentration in the material and influence of irradiation on thermal properties of cotton-silver composites were investigated with TGA and DSC methods. The proposed technique can be applied as a method for bioactive materials synthesis.
Fig.1. E.coli bacteria growth on the blank (a) and modified with silver (b) cotton samples
1. Montazer M, Behzadnia A, Pakdel E, Rahimi MK, Moghadam MB (2011) Photo Induced Silver on Nano Titanium Dioxide as an Enhanced Antimicrobial Agent for Wool. Journal of Photochemistry and Photobiology B: Biology, article in press 2. Yeo SY, Lee HJ, Jeong SH (2003) Preparation of nanocomposite fibers for permanent antibacterial effect. Journal of Materials Science 38: 2143-2147
78 Radiation Chemistry
RADIOLYTIC DECOMPOSITION OF DICLOFENAC IN WATER BY GAMMA IRRADIATION
A. Bojanowska-Czajka1, M. Trojanowicz1'2, D. Solpan3, G. Kciuk1, G. Nałęcz- Jawecki4
1 Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-19 Warsaw Poland 2 Warsaw University Department of Chemistry, Pasteura 1, 02-093 Warsaw Poland 3 Hacettepe University, Department of Chemistry, 06800 Beytepe/Ankara, Turkey 4 Warsaw University of Medicine, Department of Environmental Health Sciences, Banacha 1, 02-097 Warsaw In recent decades the radiation technologies find numerous applications in pharmaceutical science and technology. Increasing interest is observed in y-irradiation in performing sterilization of drugs. Widely investigated are the effects of irradiation on controlled drug delivery/controlled drug release systems. There is widely employed capacity of radiation to act as an initiator of cross-linking in the manufacturing and modification of a polymer drug carriers with additional advantage of reducing the microbial load of pharmaceutical preparations in the same time [1]. Radiation polymerization is used for preparation of hydrogels widely employed e.g. as sustained-release drug delivery systems and scaffolds in tissue engineering [2]. One of the most frequently detected pharmaceuticals in waters and urban wastes is diclofenac (DCF), which is 2-(2-(2,6-dichlorophenylamino)phenyl)acetic acid, used mostly in form of sodium salt or methyl ester. It is widely used in medical care as an analgestic, antiarthretic and antirheumatic compound belonging to the group of the non-steroidal anti-inflammatory drugs (NSAID). The 15% of DCF is excreted unchanged after human consumption. It is used worldwide and has a production volume estimated to be in the hundreds of tones annually [3]. It is estimated that more than 90% of DCF is eliminated from waters by photolytic degradation by natural sunlight. In this work we examined the possibility of application of ionizing radiation for decompo-sition of diclofenac residues in waters for environmental purposes. Based on the cited above literature on Advanced Oxidation Processes (AOPs) reported for the decomposition of DCF, it was assumed that the mechanism of decomposition is based mostly on oxidation, hence experiments were carried out in aerated aqueous solutions. The yield of radiolytic decomposition of DCF was examined in terms of absorbed dose range up to 4.8 kGy in solutions of initial concentrations 10 and 50 mg/L, which are similar to those examined in studies with other AOPs [4-6]. Employing two 60Co gamma sources of different activity, the lack of effect of dose-rate on decomposition of DCF was found for 50 mg/L DCF solutions. The decomposition of DCF at different initial concentration requires different absorbed dose. It was found that the complete decomposition of 10 mg/L DCF occurs at 0.8 kGy, while for 50 mg/L DCF solution at absorbed dose 4.8 kGy the 93% yield of decomposition was observed. Using pulse radiolysis system also the reaction-rate constants for diclofenac with different radical products of water radiolysis were determined. A very important supplement to this research was the monitoring of toxicity changes during irradiation using Microtox test.
1. Razem D, Katusin-Razem B, (2008) The effects of irradiation on controlled drug delivery/controlled drug release systems. Radiat. Phys. Chem. 77: 3:288-344 2. Dong LC, Hoffman AS (1990) Synthesis and application of thermally reversible heterogels for drug delivery. J. Control Release 13; 21-31 3. Buser HR, Poiger T, Müller MD, (1998) Environ. Sei. Technol, 32; 3449-3456 4. Perez-Estrada LA, Maldonado MI, Gernjak W, Agüra A, Fernandez-Alba AR, Ballesteros MM, Malato S, (2005) Catalysis Today, 101; 219-226 5. Hartmann J, Bartekls P, Mau U, Witter M, von Tümpling W, Hofmann J, Nietz-schmann E, (2008) Chemosphere, 70; 453-461 6. Rizzo L, Meric S, Kassinos D, Guida M, Russo F, Belgiorno V, (2009) Water Res. 43; 979-988
79 NUTECH-2011
SELECTION OF THE MATERIALS FOR RADIATION CROSS- LINKED CABLES
G. Przybytniak, Z. Zimek, A. Nowicki, K. Mirkowski, J. Boguski
[email protected] Institute of Nuclear Chemistry and Technology, Dorodna 16 Str., 03-195 Warszawa, Poland
Radiation technology is used as a physical method for cross-linking of some polymers, e.g. polyethylene and its copolymers. Due to excellent thermal, mechanical and chemical features so modified materials are applied in several fields of industry. Nowadays, wires and cables of cross-linked insulation are used in electronic industry, nuclear power plants, communication, railways, sea and air transportation. Radiation treatment leads to the improvement of their functional properties increasing working temperature, current rating and resistance to short circuit. The replacement of chlorine containing insulation made from poly(vinyl chloride) can eliminate the risk of hydrochloride, dioxin, phosgene and other toxic gases emission in the case of fire. The cables constructed from halogen-free substances have to meet customer requirements to the same degree as previous products. Copper conductors in cables, usually tin-coated, are covered with extruded insulation, that typically is shielded with the overall jacked. Sufficient quality of both layers might be achieved using radiation technology. Commercial application of electron accelerators in cable cross-linking started in early seventies of the last century and has developed since then gradually. The radiation processing requires high energy consumption thus it is recommended to doped the materials used for insulation with small amounts of cross-linking agents [1]. An addition of multifunctional monomers into some polymeric matrices efficiently lower absorption doses required for their cross-linking. In the reported studies, designed for radiation cross-linking commercial blends based on EVA copolymer and standard insulation polyethylene were tested to determine the optimal conditions for electron beam irradiation, using INCT electron accelerator. The studied polyethylene was modified using cross-linking activator facilitating formation of the three dimensional network, namely Rhenofit® TAC/S containing 70% triallyl cyanurate (TAIC) adsorbed on silica [2]. The mechanical and thermal properties as well as resistance to oxidizing atmosphere at elevated temperatures defined as oxidation induction time (OIT) were determined for the materials and blends irradiated with doses up to 150 kGy.
1. Goodarzian N, Shamekhi MA (2009) Chemical cross linking versus high energy electron beam cross linking ofHDPE: electrical properties study. J. Iran. Chem. Res. 2;189-194 2. Hana DH, Shina S-H, Petrov S (2004) Crosslinking and degradation of polypropylene by electron beam irradiation in the presence of trifunctional monomers. Radiation Physics and Chemistry 69;239-244
80 RadiaƟon Chemistry
WASTEWATER TREATMENT WITH MOBILE E-BEAM PLANT B. Han1, J.K. Kim1, Y. Kim1, W.G. Kang1, N. Zomme2 [email protected]
1EB TECH Co., Ltd. 550 Yongsan-dong Yuseong-gu, Daejeon, 305-500 Korea 2Pele Technology Inc., 1590 Buckeye Drive, Milpitas, Ca95035, U.S.A. One of the important environmental problems is the reuse of wastewater to industrial and agricultural applications. The use of alternative disinfectants to chlorine for the wastewater treatment has received increasing attention in recent years to treat either liquid or solids streams within wastewater treatment plants for pathogens and trace organics (TOrCs). [1] Although several technologies have come to the forefront as an alternative to chlorine (e.g., ultraviolet [UV] and hydrogen peroxide), the majority of these technologies are chemically based, with the exception of UV. An attractive physical disinfection approach is by electron beam irradiation. On-site pilot scale treatment of wastewater from municipal plant will be applied to optimize the operation parameters by a skid mounted Mobile Electron Beam Plant [2] To compare electron beam
technology with other advanced oxidation processes (AOP) such as UV and Ozone technology, in regard to costs and effects for different application cases and local situations. The objectives of this study are optimization of the EB process for minimizing energy loss and absorbed dose upon the water thickness optimization of pre-treatment and post-treatment design parameters and development of design criteria to construct commercial scale re-use plant. Also the evaluation of alternatives to address unforeseen problems resulting from the actual use of proposed treatment processes and characterization and quantification of the raw water and finished water will be conducted. 1. Vanderford BJ and Snyder SA (2006) Analysis of pharmaceuticals in water by isotope dilution liquid chromatography/tandem mass spectrometry. Environ. Sci. & Technol., 40:7312-7320 2. Kang WG, Kuk SH, Kim JK, Han B, Kang CM (2009) Shielding Design of a Mobile Electron Accelerator Using Monte Carlo Technique. Journal of Rad. Industry 3(2) 79- 85
81
NUTECH-2011
WETTABILITY OF CARBON AND SILICON CARBIDE CERAMICS INDUCED BY THEIR SURFACE ALLOYING WITH Ti, Zr AND Cu ELEMENTS USING HIGH INTENSITY PULSED PLASMA BEAMS
M. Barlak , J. Piekoszewski , Z. Werner , B. Sartowska , L. Waliś , W. Starosta , J. Kierzek K. Bocheńska1, R. Heller3, A. Kolitsch3, C. Pochrybniak1'4, E. Kowalska5
Andrzej Sołtan Institute for Nuclear Studies, Świerk/Otwock, Poland 2Institute of Nuclear Chemistry and Technology, Warsaw, Poland 3Helmholtz-Zentrum Dresden-Rossendorf, Dresden, Germany institute of Atomic Energy, 05-400 Otwock/Świerk, Poland 5Military University of Technology, Warsaw, Poland
Ceramics materials, such as: oxides, nitrides, borides, carbides and carbon are widely used in modern constructions and devices. Their advantages are: low density, high mechanical strength and corrosion resistance at high temperature and favourable performance/weight relationship. However, an application of these materials in joints or in composites with metals is very difficult, because usually the ceramics are non-wettable by liquid metals. In the present work, we used high intensity plasma pulses technique for the preparation of carbon and silicon carbide surface before the wetting process by liquid copper. The Ti, Zr and Cu plasma was applied to induce the wettability. The experiments were preceded by thermodynamical considerations. The prepared samples were investigated by sessile-drop tests, SEM observations, EPMA, GXRD analysis and RBS measurements. The results of Ti and Zr plasma modifications were beneficial and similar to each other. The measured contact angles were below 90°. The results of Cu plasma were unfavorable with contact angles close to 180°.
82 NUTECH-2011
A 2-D THERMOLUMINESCENCE DETECTOR SYSTEM BASED ON LiF:Mg, Cu, P AND CaSO4:Dy FOILS FOR QUALITY ASSURANCE IN RADIATION DOSIMETRY
M. Kłosowski1, R. Kopeć1, J. Gajewski1, D. Kabat2, K. Kisielewicz2, P. Olko1, M. Ptaszkiewicz1, T. Nowak1, M.P.R. Waligórski1'2
Mariusz.Klosowski@ifj. edu.pl
1The H. Niewodniczański Institute of Nuclear Physics Polish Academy of Science, ul. Radzikowskiego 152, 31-342 Kraków 2Centre Of Oncology Maria Sklodowska-Curie Memorial Institute Kraków Branch, Poland
Modern radiotherapy techniques, such as intensity modulated radiation therapy (IMRT), stereotactic radiotherapy or proton therapy, together with other newer radiation delivery platforms such as CyberKnife® or TomoTherapy®, require complex dosimetry QA and verification procedures. Individual verification of the planned dose distribution is typically performed by simulating the treatment plan in a phantom and comparing the measured and planned dose distributions. In clinical practice, high resolution two-dimensional (2-D) detectors, such as as X-ray photographic emulsion or radio-chromic films are largely used for verification of dosimetry [1]. Thermoluminescent detectors (TLDs) are also used in clinical dosimetry but mostly as point detectors for in-phantom measurements of dose distribution [2]. 2-D TL sheets, obtained from a mixture of highly-sensitive phosphors, LiF:Mg,Cu,P TL powder or CaSO4:Dy were developed at the Institute of Nuclear Physics in Krakow together with a planar readers equipped with Charge Coupled Devices (CCD) camera [3]. Due to good linearity of dose response (range 20 cGy - 20 Gy) (Kisielewicz et al., 2010) and spatial resolution better than 1 mm, our 2-D planar TLD system can potentially complement existing 2-D techniques. Moreover, detector reusability makes the dosimetric verification procedure less costly. In this work we present potential areas of applicability of our 2D TL systems, such as IMRT dose verification, proton beam monitoring and possible verification of static/dynamic exposition for individual dosimetry monitoring.
1. Winkler, P., Zurl, B., Guss, H., Kindl, P., Stuecklschweiger, G., 2005. Performance analysis of a film dosimetric quality assurance procedure for IMRT with regard to the employment of quantitative evaluation method. Phys. Med. Biol. 50, 643-54 2. Bilski, P., Waligorski, M.P.R., Budzanowski, M., Ochab, E., Olko, P., 2002. Miniature thermoluminescent detectors for dosimetry in radiotherapy. Radiat. Prot. Dosim. 101, 473-476 3. Marczewska, B., Bilski, P., Czopyk, Ł, Olko, P., Waligórski, M.P.R., Zapotoczny, S., 2006. Two-dimensional thermoluminescence dosimetry using planar detectors and a TL reader with CCD camera readout. Radiat. Prot. Dosim. 120, 129-132. 4. Kisielewicz, K, Swiebocka, J., Czopyk, Ł, Kłosowski, M., Lesiak, J., Byrski, E., Kabat, D., Wawrzak, M., Śladowska, A., Dziecichowicz, A., Olko, P., Waligórski, M.P.R, 2010. Dosimetric properties of TL foils based on LiF:Mg,Cu,P (MCP-N) phosphors for clinical applications. Radiat. Meas. 45, 716-718.
84 Dosimetry and Radiation Protection
A COMPARISION OF METHODOLOGY OF DOSE CALCULATION METHODS FOR ASYMMETRIC FIELDS IN NUCLEAR TECHNOLOGIES
S.H.Masoumi
PhD Student in medical physics, Tarbiat Modares University
Several methods have been developed for the dosimetry of asymmetric radiation fields formed by independently moving collimator jaws. Three of these methods were utilized for the calculation for asymmetric fieldinnuclear technologiesdose profiles. All three methods use for output factors(OFS) and percent depth doses (%DD) or tissue maximum ratios (TMR) of symmetric fields. In the first methods, calculation of the off -center ratio (OCR) of the asymmetric field is based on the symmetric field from which the asymmetric field is originated, by setting one jaw in an asymmetric position. In the second method the OCR of the symmetric field is used for the off-center ratio asymmetric (OCRa) are calculation of the asymmetric field of the small size; whereas the third method dose not allow for the asymmetric OCR calculation. The result obtained for the 9MV photon beam of a Nepton 10PC linear accelerator. %DD, isodose and point dose measurements where done for asymmetric fields of 6*6 with offset of 3cm, 10*10 with offset =5cm, 15*15 with offset= 7.5cm, using Scanditronix radiation fields detectors and RF300 plus water phantom. The results show that although calculations done using the Kwa method are less accurate when compared to Khan and Loshek method of calculations, this method uses fewer factors for dose calculation and the factors are easier to calculate. Khans method of calculations dose results in very accurate result when calculating dose in the center average and large fields (an accuracy 0.5%). Compared to the other two methods, Losheks method is the most accurate method when the dose in the center is compared to the dose in the edge of the field.
1. William Kwa, Richard O. Kornelsen, Richard W. Harrison, and Ellen El-Khatib,Dosimetry for asymmetric x-ray fields Med. Phys. 21, 1599 (2007); doi:10.1118/1.597260 (6pages) 2. N. Fournier-Bidoz, Y. Kirova, F. Campana, J. El Barouky, S. Zefkili, Technique alternatives for breast radiation oncology: Conventional radiation therapy to tomotherapy, J Med Phys. 2009 Jul-Sep; 34(3): 149-152. 3. Khaled Abdel-Hakim, Tetsuo Nishimura, MichikatsuTakaih, Dosimetric Assessment of the Field Abutment Region in Head and Neck Treatments Using a Multileaf Collimator,Strahlentherapie und Onkologie(2010) Volume 179, Number 5, 312-319, DOI: 10.1007/s00066-003-1024-1
85 NUTECH-2011
A NOVEL MONITOR OF NEUTRON - GAMMA RADIATION DOWN TO ENVIRONMENTAL LEVELS
S.Pszona
[email protected] The Andrzej Soltan Institute for Nuclear Studies,04-500 Otwock, Swierk, Poland
The new method for measuring ambient dose equivalent in mixed neutron - gamma field has been invented. It has been shown that the moderator based technique, up to now used for neutron monitoring only, can be adjusted for monitoring both neutron and gamma radiation. It has been shown that fluence response to neurons and gamma of a device consisting of a He-3 proportional counter inside a 203 mm diameter poltyhtene sphere are almost identical. model device based on above described detector assembly has been costructed and tested. A modified proportional counter filled with He-3 to 160 kPa by adding a guard ring made by ZDAJ IPJ has been applied. Such modification resulted in reduction of the electronic noises of used proportional counter, very substantially, allowing a safe registration of pulses from gamma radiation. A model of an instrument capable of preamplifying signals from the detector, shaping them for accurate peak detection, digitizing peak amplitude and maintaining a 256 channel histogramme on 64X256 LCD graphic screen. The histogramme, together with other parameters such as HV bias of a detector, astronomical time and measurement time can be stored on a non-volatile SRAM card. Steering and control functions are performed by a H8/532 microprocessor. The data processing function allows viewing of the stored data as number of counts per channel or as a pulse height spectrum; finally sending selected data, stored in 120 bank SRAM card, to a printer or PC computer by an RS-232 communication interface. The measuring time can be selected in 1 s steps up to 24 h and up to 120 cycles. The model device called" environmental monitoring station" has been calibrated in Pu-Be source as well as by a gamma Cs-137 source. The background of the model device, especially at the neutron channels, has been estimated from measuring runs at low level laboratory at Asse salt mine. The results of bacground measurements in an equivalent in nSv units as well as the calibration results as well as the other performances of the device will be given.
86 Dosimetry and Radiation Protection
ALANINE DOSIMETRY OF 60 MeV PROTON BEAM AT IFJ PAN - PRELIMINARY RESULTS
B. Michalec1, G. Mierzwińska, U. Sowa1, T. Nowak1, J. Swakoń1
Barbara.Michalec@ifj. edu.pl lrThe Henryk Niewodniczański Institute of Nuclear Physics Polish Academy of Sciences , ul. Radzikowskiego 152, 31-342 Kraków, Poland
The Electron Paramagnetic Resonance spectrometry (EPR) of the amino acid L-a alanine CH3-CH(NH2)-COOH used as a radiation sensitive material has been known as a dosimetry method for over 40 years. It has been applied mainly for high dose measurements in food preservation and medical equipment sterilization [1]. In the last years EPR/alanine dosimetry has gained attention as a potential tool for proton and ion dosimetry in radiotherapy because it is characterized by a composition quite similar to that of a tissue, linear response to dose and high sensitivity to dose therapy levels [2]. Nowadays, the clinical proton dosimetry is mainly based on active dosimeters, first of all on ionization chambers. The extended uncertainty for the primary dosimetry in radiotherapy should not exceed 5% which is extremely difficult to fulfill for a passive, integrating dosimeters. These, less precise, types of dosimeters (e.g. radiochromic films, TLDs or alanine detectors) are then broadly used for the quality control of the therapeutic beams. Among them alanine, as the less LET dependent, is considered as one of the most promising passive dosimeters, however implementation of the EPR/alanine spectrometry as a tool for proton dosimetry requires the precise examination and assessment of alanine efficiency for particular BP regions of the variable LET values as well as taking into consideration other factors affecting the dose estimation characteristic for the given facility and read out system. The Institute of Nuclear Physics Polish Academy of Sciences (IFJ PAN) carries out a project EPR/alanine dosimetry for radioterapeutic ion beams, realized in the frames of PARENT- BRIDGE Programme introduced by the Foundation for Polish Science. Its general objective is developing methods and procedures, based on EPR spectrometry of alanine, for dosimetry of radiotherapy ion beams, particularly for proton beams operating at IFJ PAN; namely 60 MeV proton beam from AIC-144 cyclotron delivered to the facility for ocular tumor irradiation as well as for 235 MeV proton beam from Proteus 235, which will be available at Cyclotron Center Bronowice since 2013. The paper will show the preliminary results of this work - determination of linearity of proton and Co-60 dose response, dose response of alanine in modulated beam (spread out Bragg peak region) and alanine relative effectiveness to protons with respect to Co-60 as a function of proton energy.
1. Regulla DF, Deffner U (1982) Dosimetry by ESR spectroscopy of alanine. Int. J. Radiat. Isot. 33:1101 2. Onori S, Bortolin E, Calicchia A, Carosi A, De Angelis C, Grande S (2006) Use of commercial alanine and TL dosemeters for dosimetry intercomparisons among italian radiotherapy center. Radiat Prot Dosim. 120(1-4):226-229
87 NUTECH-2011
APPLICATION OF LiF:Mg,Cu,P (MCP-N) THERMOLUMINESCENT DETECTORS FOR EXPERIMENTAL VERIFICATION OF RADIAL DOSE DISTRIBUTION MODELS
W. Gieszczyk, P. Olko, P. Bilski, L. Grzanka, B. Obryk
Institute of Nuclear Physics Polish Academy of Sciences (IFJ PAN), Cracow, Poland
Radial distribution of dose, D(r), around cores of heavy ions tracks, is a base concept of several radiobiological models of cell survival. Until now, the D(r) models were mainly used to predict biological effects and response of physical detectors on doses of ionizing radiation. Several models of D(r) predict that at the radial distances closer than 1 nm from the core of heavy ion track, local dose can reach values as high as 106 Gy. These values have never been verified experimentally. Furthermore, no one has ever worked out a method of experimental verification of D(r) models using thermoluminescent (TL) detectors, because their response, on ionising radiation doses, saturates typically at dose of 103 Gy. This work proposes an innovative method of experimental verification of radial dose distribution models using LiF:Mg,Cu,P (MCP-N) thermoluminescent detectors (TLD), which show high temperature TL glow peak structure between 350oC and 550oC after exposure to doses of gamma-rays as high as 1 MGy. MCP-N detectors were irradiated with Am - 241 a - particles with fluences 107 - 1011 particles/cm2. All evaluated TL glow curves were deconvoluted into N single peaks using GlowFit code. For each TL peak i the equation connecting an intensity, I, of TL signals from alpha and gamma irradiations with dose-frequency function, f(D), was written:
ra=\i;(D)-f(D)dD, i = \..,N
Having measured IJ and //, the f(D) was unfolded and converted to D(r). The parametric unfolding and the unfolding based on SAND-II iterative algorism were applied. The first results show that the D(r) decreases as 1/r2, as predicted by analytical models. For the intermediate radius range the solutions do not fit to analytical models. These discrepancies will be discussed in the paper. The method gives the first experimental approach to determine the radial dose distribution around the path of heavy charged particles in LiF detectors.
1. Waligórski MPR, Hamm RN, Katz R (1986) The radial distribution of dose around the path of a heavy ion in water. Radiat. Meas. 11, 309-319. 2. Wingate CL, Baum JW (1976) Measured radial distributions of dose and LET for alpha and proton beams in hydrogen and tissue-equivalent gas. Radiat. Res. 65, 1-19. 3. Bilski P, Obryk B, Olko P, Mandowska E, Mandowski A, Kim JL (2008) Characteristics of LiF:Mg,Cu,P thermoluminescence at ultra-high dose range. Radiat. Meas. 43 (2008) 315-318 4. Puchalska M, Bilski P (2006) GlowFit-a New Tool for Thermoluminescence GlowCurve deconvolution. Radiat. Meas. 41, 659-664
88 Dosimetry and Radiation Protection
APPLICATION OF RETROSPECTIVE BIOLOGICAL DOSIMETRY WITH DICENTRICS AND FISH TECHNIQUES FOR ACCIDENTAL EXPOSURES TO RADIATION IN POLAND (1996-2010)
A. Cebulska-Wasilewska1, J. Miszczyk1, J. Swakoń2
1 Institute of Nuclear Physics PAN, Department of Radiation and Environmental Biology, Radzikowskiego 152, Krakow, Poland 2 Institute of Nuclear Physics PAN, Proton Radiotheraphy Group, Radzikowskiego 152, Krakow, Poland
In Poland in the period between 1996 till 2010, blood samples from persons that were suspected of accidental exposure to ionizing radiation were investigated in our laboratory for the purpose of the retrospective biological dosimetry. To estimate the potential health risk of exposure, lymphocytes from their blood have been analyzed for the presence of chromosomal aberrations particularly dicentrics and rings, or chromatid exchanges with classic cytogenetics or translocations with FISH techniques. Results were compared to the dose response curves for dicentrics, detected in our laboratory after induction in human lymphocytes by in vitro irradiation with various LET radiation from different therapeutic sources (EC Petten, BNL medical reactor, Cf-252, Co-60, protons 60 MeV-AIC-144 Krakow, X-rays), or translocations with FISH techniques (Cf-252, protons 60 MeV, X-rays). Dose response curve for gamma radiation, evaluated under the same laboratory and culturing procedure conditions, was applied to evaluate the absorbed dose. The frequency of dicentrics and rings, detected in lymphocytes from some of the investigated persons, have indicated excessive dose, several times about the permissible annual limit. Similar results were found for translocations with FISH techniques, as well as with DNA repair competence assay. Results of dicentrics are discussed in terms of relative health risk from cancer, based on survival rate estimated in our follow up studies in population monitored with chromosomal type of aberrations during the period 1980-2003.
Acknowledgments: Research was partially supported by grant: "Studies on individual susceptibility to radiation- induced damages, DNA repair and instability of human chromosomes in lymphocytes of patients diagnosed or treated with radioiodine" (0296/B/P01/2008/3 5) and "Estimation of individual radiosensitivity of patients with prostate cancer and benign prostatic hyperplasia with application of FISH technique " (2520/B/P01/2010/39).
89 NUTECH-2011
CHARACTERISTICS AND PROPERTIES OF SOLID STATE DETECTORS IN A 60 MeV PROTON BEAM
U. Sowa, T. Nowak, B. Michalec, G. Mierzwińska, J. Swakoń, P. Olko
The Henryk Niewodniczański Institute of Nuclear Physics Polish Academy of Sciences, ul. Radzikowskiego 152, 31-342 Krakow, Poland
Solid state detectors can be useful for the studying of dosimetric characterization and quality control of radiotherapy proton beams. Diamond is considered as a very promising material for use as radiation dosimeter. Small size, high sensitivity to the radiation, fast response and tissue equivalence make it possible to use it in charged particle monitoring. Although the calibration method is unresolved problem. The TRS-398 IAEA code of practice provide recommendations for ionization chamber dosimetry, based on calibrations in Co-60 beam in terms of absorbed dose to water to use it in dosimetry of the proton beams. This protocol provide the calculated values for chamber specific factors, kg, as a function of beam quality index, Rres for cylindrical and plane-parallel ionization chambers (1). To determine the empirical values of kq for diamond detectors authors adapt the method that takes into account the LET dependence of this detectors (2). The research of suitability of diamond detectors in proton dosimetry field was presented, especially dose and dose rate dependence and relative dose distribution. Each aspects were compared with the Markus ionization chamber. The results were compared also with the PTW silicon diode. Irradiations were performed in a water phantom with the 60 MeV ocular therapy beam produced by an isochronous cyclotron AIC-144 localized at the Institute of Nuclear Physics in Krakow. The maximum proton range measured in water phantom installed at the isocenter of the facility is about 28.3 mm and the dose rate values can be changed from 0.1 to 0.7 Gy/s. This dosimetric measurements are the first step in further research of using the other type of detectors (alanine detector, thermoluminescent detector) in proton beam dosimetry.
Acknowledgments: This work was supported by the Foundation for Polish Science "POMOST Programme" co- financed by the European Regional Development Fund (Innovative Economy Operational Programme 2007-2013).
1. IAEA (International Atomic Energy Agency), Absorbed Dose Determination in External Beam Radiotherapy: An international Code of Practice for Dosimetry based on Standards of Absorbed Dose to Water, IAEA TRS No. 398, Vienna (2000) 2. Fidanzio A, Azario L, De Angelis C, Pacilio M, Onori S, Kacperek A, Piermattei A, (2002) A correction method for diamond detector signal dependence with proton energy. Med. Phys. 29(5): 669-695
90 Dosimetry and Radiation Protection
COMBINED TL AND OSL READOUT OF LiF DETECTORS
B. Marczewska1, P. Bilski1, E. Mandowska2, A. Mandowski2
Barbara.Marczewska@ifj. edu.pl
1 Institute of Nuclear Physics PAN, ul. Radzikowskiego 152, 31-34 Kraków, Poland 2 Jan Długosz University, ul. Armii Krajowej 13/15, 42-200 Częstochowa, Poland
Thermoluminescence (TL) and Optically Stimulated Luminescence (OSL) are the very well known passive methods of dose measurements applied in radiation protection. Lithium fluoride (LiF) detectors, routinely used in TL dosimetry, can be used as luminescent detectors stimulated by a light during the readout. The emission curve obtained from an irradiated LiF detector and excited by a blue light (460nm) consists of two peaks at the wave lengths of 535nm and 650nm [1]. Oster has also reported [2] that TL processes and light induced effects connected with other types of traps are independent and not related to each other. It gives the possibility to get the discrimination of radiation field due to different efficiency of LiF detector to various modality of radiation if the combined TL and OSL readout even of one LiF detector is performed. The preliminary experiments on excitation and emission spectra measurements of LiF detectors were done at Jan Długosz University in Częstochowa. Luminescence signal of LiF detectors was investigated as a function of irradiation dose/fluence for protons, gamma rays and alpha particles. The results showed very low efficiency of OSL effect of LiF irradiated with gamma rays or protons in opposite to high efficiency of OSL after alpha particle radiation. The amplitude of emission light of LiF detector irradiated by alpha particles and excited by a blue light is constant during successive measurements what indicate that instead of OSL the photoluminescence effect is rather present. LiF: Mg,Ti detectors irradiated with protons, alpha particles and gamma rays were then read out in a special newly constructed portable reader, equipped with blue light diodes, excitation and emission interference filters and a Hamamatsu photomultiplier. The first results showed the high difference between radiophotoluminescence and thermoluminescence effects of investigated detectors, which is dependent on the modality of the radiation. The further investigation will be done to develop a method of discrimination of radiation in mixed-radiation fields by means of standard LiF detectors based on combined TL/OSL measurements.
1. Oster L., Horowitz Y.S., and Podpalov L. (2008) OSL and TL in LiF:Mg, Ti following alpha particle and beta ray irradiation: application to mixed-field radiation dosimetry, Radiat. Prot. Dosim. 128; 261- 265 2. Oster L., Horowitz Y.S., and Podpalov L. (2010) OSL and TL in TLD-100 following alpha and beta irradiation: application to mixed-field radiation dosimetry, Radiat. Meas. 45; 1130-1133
91 NUTECH-2011
COMPARISON OF APPLICATIONS OF GENE MUTATION ASSAY IN Trad-SH CELLS FOR MONITORING AMBIENT AIR GENOTOXICITY AFTER CHERNOBYL AND FUKUSHIMA NUCLEAR POWER PLANT ACCIDENTS
A. Panek, J. Miszczyk, A. Cebulska-Wasilewska,
Institute of Nuclear Physics PAN, Department of Radiation and Environmental Biology, Radzikowskiego 152, Krakow, Poland
In order to define major contamination, its genotoxic effectiveness and to realize genetic or carcinogenic hazards, studies are necessary which yield new information on mutagenicity of complex mixtures containing potential genotoxic pollutants. The technique for screening gene mutation frequency in somatic cells of the Tradescantia stamen-hairs (Trad-SH assay), have been developed many years ago specifically for radiobiological studies. Tradescantia is one of the most radiosensitive plant known. Its extremely high radiosensitivity of its hybrid clones is followed by very high sensitivity to chemical mutagens as well. This facts make Trad-SH assay particularly suitable for the environmental studies and for the detection of ambient air genotoxicity. Results of applications of the bio-indicator for in situ monitoring gentoxicity of the ambient air pollution including ionizing radiation from Chernobyl Nuclear Power Plant accident are compared to recent data from monitoring the ambient air quality in the Krakow and surroundings. Following the Chernobyl accident studies were performed initially as monitoring of mutagenicity of ambient air in the period since April 29th till June 3rd 1986. Significant increase of gene mutation frequency was reported, associated with a strong expression of toxic effects. In general, mutation frequency increase due to Chernobyl fallout was corresponding to fluctuation of radioactivity in the air reported from physical measures, and to published reports about increase in chromosome aberration levels. One year later studies were repeated and mutation frequency tested at site the same as year before confirming decrease of mutation rate to the control level. Monitoring of genotoxicity of ambient air was also performed around various pollution sources (petroleum plant, heavy traffic lines) and detected gene mutation frequencies levels corresponded well either with concentrations of chemicals in ambient air or with distance from the source of pollution. Recently since 11th of March 2011, bio-indicating plants are exposed to ambient air at four different sites at the region of Krakow, and continuous screening is performed. Results will be discussed with the concern of the possible association with the physical measures of ambient air radioactivity.
Acknowledgments: Research was partially supported by grant: "Studies on individual susceptibility to radiation- induced damages, DNA repair and instability of human chromosomes in lymphocytes of patients diagnosed or treated with radioiodine" (0296/B/P01/2008/3 5) and " Estimation of individual radiosensitivity of patients with prostate cancer and benign prostatic hyperplasia with application of FISH technique "(2520/B/P01/2010/39).
92 Dosimetry and Radiation Protection
DEDICATED COMPUTER SOFTWARE TO RADIATON DOSE OPTIMIZATION FOR STAFF PERFORMING NUCLEAR MEDICINE PROCEDURES
J. Kosek, K. Matusiak
AGH University of Science and Technology, Krakow, Poland
Nuclear medicine use imaging techniques to observe proper or pathological physiology of various human organs. To make this visualization possible, radioactive isotopes connected with ligands - known as radiopharmaceutics, have to be administered to the patient. Each application of radiopharmaceutic is proceed by its suitable preparation. As a result, medical staff working with radioisotopes is continuously exposed on ionizing radiation [1]. Moreover, the major value of absorbed dose was observed for the staff working in "hot labs". Dose overestimation derived from the radiopharmaceutic's preparation protocol. The everyday procedure contains a few (for advanced employee) or many (for beginner employee) volume and radioactivity verifications before final handing over to the application. What is more, the time of radiation influence on the workers extended and as a result value of absorbed radiation dose is increasing. Presented software for radiation dose optimization minimize the number of preparation steps through real time estimation of dynamic changes in time for radiopharmaceutics radioactivity and volume. As a result, the absorbed dose level was significantly reduced. It was verified by simultaneously TLD measurements. Presented software was dedicated for Microsoft Windows system family. It was based on .NET 3.5 framework and SQL Server Compact database. It is planned to provide a version for Linux system as well. Integration between presented application and existing in hospitals database systems is possible. User login and password is provided to control preparation process. Set up of each radiopharmaceutic preparation is connected with current user. There were informations about generator type and maximum of radioactivity that was eluated during one day. For more than one eluation, corrector factor are available. All necessary data connected with patient such as name, surname, ID number, recommended radiopharmaceutic with radioactivity etc. were taken into consideration. To make one's work easier list of previous, current and later application were displayed with proper colours i.e. green for carry out, red for waiting to realization. Beside the main functionality daily and based on history reports were offered.
Key words: nuclear medicine, computer software, radiation dose estimation, medical staff 1. Stabin MG (2008) Radiation Protection and Dosimetry An Introduction to Health Physics Springer 2008, New York, USA
93 NUTECH-2011
DEVELOPMENT OF CALIBRATION PROCEDURE FOR DOSE CALIBRATORS WITH USE OF THE EFFICIENCY CURVES
T. Dziel, A. Patocka, A. Muklanowicz
Radioisotope Centre, Institute of Atomic Energy POL ATOM, 05-400 Otwock-Świerk, POLAND
According to the regulation of the Polish Minister of Health dated February 18th, 2011 about conditions for safe use of ionizing radiation exposure for all types of medical work nuclear medicine departments are required to perform once a year a specialized test of accuracy of dose calibrators with use of the certified standard sources of each radionuclide used in laboratory [1]. Laboratory of Radioactivity Standards (LRS) in Radioisotope Centre (RC), IAE POLATOM provides calibration services for dose calibrators with ionization chambers on the basis of the isotopes in the continued production in the RC (among others 99mTc, 131I, 90Y, 89Sr). In subsequent years there is a steady increase in the number of orders. At the same time the queries are repeating on the possibility of calibration with the use of isotopes with limited availability (or its absence) in the Centre (i.e. 67Ga, 153Sm, 169Er). The use of these isotopes in the way of the currently used procedure is uneconomic both from the point of view of the customer as well as LRS. Development of a new calibration procedure for determining the calibration coefficients based on measurement of efficiency curves of ionization chambers allows the LRS to expand a service offer and to meet the growing demands of the customers. Efficiency curves of the ionization chambers used in the LRS and for the commercial dose calibrator, CAPINTEC CRC-15beta, were examined. Various analytical form of the curves were studied to minimize uncertainty related to the quality of the fit. At the same time the attention was paid to make elaborated procedure simply and fast enough to meet customer's requirements. Studies have shown that the method gave satisfactory results for radionuclides with a predominance of emission lines with energies above 100 keV. Differences between the experimental and calculated coefficients do not exceed 3%. For radionuclides with low- emission energy direct calibration remains as the best solution. Also in the case of pure beta emitters it is need to use the traditional method as a part of defining the beta component of the calculated efficiency curve. Due to the reduction in the cost of the application of the method, it was decided to complete a set of standard sources of used radionuclides with sufficiently long half-lives, which will be allowing their use in the few subsequent calibrations. The set included: 57Co, 60Co, 133Ba, 137Cs and 152Eu.
1. Journal of Laws of Republic of Poland (2011), No. 51, pos. 265, p. 3229-3280
94 Dosimetry and Radiation Protection
ENERGY AND DOSE RESPONSE OF 2-D THERMOLUMINESCENT FOILS: TYPE LiF:Mg,Cu,P AND CaSO4 FOR RADIOLOGY PURPOSES R. Kopeć and M. Kłosowski
Institute of Nuclear Physics Polish Academy of Sciences (IFJ PAN), ul. Radzikowskiego 152, PL 31-342 Krakow, Poland
A measurement of thermoluminescence consists of increasing of temperature of TL material in a controlled way with simultaneous measurement of emitted light. Measurements of light intensity were up to now realized almost entirely with use of photomultipliers (PM). In 2005 thermoluminescent (TL) reader with CCD camera was developed at the Institute of Nuclear Physics of the Polish Academy of Sciences (IFJ) Kraków. The ultra sensitive 12-bit monochromatic CCD camera was applied instead of PM. Two-dimensional (2-D) thermoluminescence (TL) dosimetry system was developed and tested by evaluating 2-D dose distributions around radioactive sources for different purposes e.g.: radiotherapy and individual dosimetry (distinguishing between static and dynamic exposure) [1].
In this study thermoluminescence foils based on LiF:Mg,Cu,P and CaSO4:Dy phosphors were calibrated and tested for 2-D dosimetry submitted to different radiological procedures. Irradiations were performed at the calibration laboratory. The dosemeters were exposed using a narrow or wide beams of X-rays of mean energies from 33 keV (N-40 quality) till 160 keV (N-200 quality). 2-D dose distributions were evaluated using a prototype planar clinical reader with heating plate 20 cm x20 cm, equipped with a 16 bit Charge Coupled Devices (CCD) camera, with a resolution of 1024 x 1024 pixels. It is able to register maximum 65000 counts in each pixel. 2D TL technology will be further improved to obtain doses obtained by patients during interventional radiology procedures.
1. Olko P., Czopyk L., Kłosowski M, Waligórski M.P.R. (2008) Thermoluminescence dosimetry using TL-readers equipped with CCD cameras, Radiat. Meas., 43; 864-869
95 NUTECH-2011
EVALUATION OF ABSORBED DOSE DISTRIBUTION IN "WIERZCHOWSKA GÓRNA" LIMESTONE CAVE
B. Karabin and A. Jung
AGH University of Science and Technology, Department of Medical Physics and Biophysics, Krakow, Poland
The risks to human health from radioactivity in underground caves, especially in Polish Jura region, have been poorly documented unlike in other workplaces or even domestic environments. Limestone rocks are usually less radioactive than granite rocks, as they emanate less radon which is responsible of about half of the annual average effective dose received by the human due to natural sources of radiation, however they are more porous, what causes that radon may emanate from deeper layers. Moreover, a higher level of radioactivity may characterize some areas of the caves [1-4]. The purpose of this study was to estimate absorbed dose in the biggest limestone cave of Polish Jurassic Highland called "Wierzchowska Górna", which has not received adequate attention so far. Measurements of spatial and quarterly radioactivity were carried out in order to assess the radiological hazard and its possible changes influenced by seasonal variations. Thermoluminescent detector MCP-N (Li:Mg,Cu,P) produced by the Institute of Nuclear Physics "PAN" in Krakow were applied for measurements. All detectors were annealed, calibrated and protected against humidity before each use, and then left in twelve places of the cave for one month in a preliminary study, and twice for three months to observe possible seasonal variation. Detectors were placed always in the same places, in and behind the tourist route. Depending on seasonal changes and ventilation the value of absorbed dose was in a range of 0.012.0.111 mGy per quarter. Lower doses were measured closer to the cave entrance and the highest dose in the "Primitive man room". Doses measured in winter were comparable, or lower depending on location, to those detected in summer. The map of absorbed dose distribution was obtained for the period July 2010-February 2011, a following measurement is under way. Measurements provided so far shown a low and not dangerous for tour guides employed at the cave level of measured absorbed dose. However, the applied method may be useful to determine seasonal changes of radioactivity for an improved description of the ventilation processes inside the cave.
1. Gillmore GK, Philips PS, Denman AR, Gilbertson DD (2002) Radon in the Creswell Crags Permian limestone caves. Journal of Environmental Radioactivity 62: 165-179 2. Oh YH, Kim G (2011) Factors controlling the air ventilation of a limestone cave revealed by 222Rn and 220Rn tracers. Geosciences Journal 15; 1:115-119 3. Sperrin M, Denman T, Phillips PS (2000) Estimating the dose from radon to recreational cave users in the Mendips, UK. Environmental Radioactivity 49: 235-240 4. Sainz C, Quindos LS, Fuente I, Nicolas J, Quindos L (2007) Analysis of the main factors affecting the evaluation of the radon dose in workplaces: The case of tourist caves. Journal of Hazardous Materials 145: 368-371
96 Dosimetry and Radiation Protection
MEASUREMENT AND CALCULATION OF EXISTING GAMMA DOSE IN ENVIRONMENT DURING THE DETERMINATION OF NITROGEN CONTENT OF EXPLOSIVE MATERIALS BY PGNAA METHOD USING MCNPX CODE
M.N. Nasrabadi1 , S. Omidi
Department of Nuclear Engineering, Faculty of Advanced Sciences & Technologies, University of Isfahan, Isfahan 81746-73441, Iran
With increasing the use of radiation in medicine, industry, and laboratories, requirements for safe and optimal use of radiation is really felt. One of the cases, in which people expose to radiation, is during the detection of explosive materials by PGNAA method. So external dosimetry is necessary for workplaces where this method is used for detecting explosives. In this study, the Monte Carlo simulation program, MCNPX has been used for simulating gamma dose in environment during the detection of explosive materials by PGNAA method. For validating this simulation, dosimetry was done in practice with a Geiger Mueller dosimeter, and indicates that the simulated and measured data are in good agreement with each other. So this study demonstrates that MCNPX code can be effective and useful for simulating gamma dose in various environments.
1. Nair AGC, et al., (2004) Analysis of alloys by prompt gamma-ray neutron activation. Nuclear Instruments and Methods in Physics Research A 516; 143-148 2. Nasrabadi MN, Mohammadi A, Jalali M (2009) Gamma self- shielding correction factors calculation for aqueous bulk sample analysis by PGNAA technique. Applied Radiation and Isotopes 67; 1208-1212 3. Im HJ, et al., (2009) Classification of materials for explosives from prompt gamma spectra by using principal component analysis. Applied Radiation and Isotopes 67;1458- 1462 4. Chichester DL, Empey E, (2004) Measurement of Nitrogen in the body using a commercial PGNAA system-phantom experiments. Applied Radiation and Isotopes 60; 55-61 5. Clifford ETH, et al., (2007) A militarily fielded thermal neutron activation sensor for landmine detection. Nuclear Instruments and Methods in Physics Research A 579; 418- 425
97 NUTECH-2011
NANODOSIMETRY - A NEW TOOL FOR DESCRIPTION OF RADIATION ACTION ON THE NANOSTRUCTURES
S.Pszona and A. Bantsar
The Andrzej Soltan Institute for Nuclear Studies,04-500 Otwock, Swierk, Poland
An overview of the latest developments in experimental nanodosimetry based on the set up called "Jet Counter",JC, is presented. JC is an unique nanodosimetric device used for experimental studying the formation of ionization clusters by single charged particle crossing a nanometric volume. Here it was assumed that the frequency distributions of ionization clusters formed in the specified gaseous nanometric sized sites are equivalent to such distributions at short segment of DNA (or more complicated forms i.e. nucleosome ,chromatine fibre ). A gaseous nano-site is scaled to liquid water equivalent nano- site(for nitrogen and propane ). Therefore a frequency distributions of ionization cluster within a specified nano site in nitrogen by a charged particle can be assumed as basic stochastic descriptor of radiation action on DNA structures. From the point of view of radiation damage to DNA, the frequency distribution of forming cluster size equal to 1 and the cumulative frequency of cluster size equal or higher than 2 are of particular importance for interpreting of single and double strand brakes in DNA. Based on frequency distributions of ionization clusters the two other nonstochastic descriptors are derived, namely: the first moment of frequency distribution as well as the cumulant of a distribution. The example of the measured frequency distributions for alpha particles and low energy electrons for DNA equivalent nanosites will be presented. The recent results of ionization ckluster distribution for Auger electrons of I-125 are included. The presented method and device can be used as a tool for elucidating the radiation damage patterns connected with targeted radiotherapy with a specific radionuclide such as At-211 and I-125. The method and device is a first approach to fluence based measuring system, which may be applied also for radiation protection purposes.
98 Dosimetry and Radiation Protection
ON THE QUALITY OF RADIATION PROTECTION IN SELECTED NUCLEAR MEDICINE DEPARTMENTS PERFORMING SCINTIGRAPHY AND PET-CT IN POLAND
R. Kopeć1, M. Budzanowski1, A. Budzyńska4'5, R. Czepczyński2'3, E. Dziuk5, M. Dziuk4, J. Sowiński2, A. Wyszomirska2'3, M. P. R. Waligórski1
Renata.Kopec@ ifj.edu.pl
1 Institute of Nuclear Physics Polish Academy of Sciences (IFJ PAN), ul. Radzikowskiego 152, PL 31-342 Kraków, Poland 2 Department of Endocrinology and Department of Surgery, Poznan University of Medical Sciences, Poznań, Poland 3 PET/CT Centre, Euromedic Diagnostics, Poznań, Poland 4 Mazovian PET/CT Centre, Euromedic Diagnostics, Warsaw, Poland 5 Department of Nuclear Medicine, Military Institute of Medicine, Warsaw, Poland
The Laboratory of Individual and Environmental Dosimetry (LADIS) performs measurements of occupational doses for over 4 000 medical, technical and research institutions in Poland. From the statistics gathered at LADIS, nuclear medicine was found to be the main source of occupational doses for medical staff in Poland. Of Polish radiation workers in nuclear medicine only some 65% of individual whole body doses and some 55% of individual extremity doses (measured using ring dosemeters) remain within natural radiation background levels, while for conventional radiology this level is observed for about 90% of those monitored. We analysed individual doses acquired by medical staff performing typical nuclear medicine procedures involvng preparation and application of beta, gamma and positron emitters (I-131, Y-90, Sm-153, Tc-99m, Lu-177, F-18) in two diagnostic scintigraphy laboratories and two PET-CT centers. The doses to medical staff were measured in terms of personal whole body dose equivalent Hp(10), doses to extremities (hands) - in terms of Hp(0.07), and eye-lens doses - in terms of Hp(3). MCP-N (LiF:Mg,Cu,P) and MTS-N (LiF:Mg,Ti) thermoluminescence detectors were used in badges and ring dosemeters. Eye-lens doses were evaluated in terms of Hp(3) with a new eye-lens dosemeter developed by Radcard within the ORAMED project. Individual doses acquired during standard nuclear medicine procedures were compared. The structure of the medical staff (technical staff, nurses, doctors) was established according to the procedures performed by them (preparation or administration of radiopharmaceuticals, patient handling). This structure allowed us to categorize individual exposure data with respect to specified fields of personnel activity. We believe that individual exposure data gathered in real operating conditions of nuclear medicine laboratories will enable us to improve the quality of radiation protection in nuclear medicine departments in Poland.
99 NUTECH-2011
RADIATION RISK CAUSED BY ENHANCED NATURAL RADIOACTIVITY
B. Michalik
h.michalikagig. eu
Laboratorium Radiometrii, Główny Instytut Górnictwa, Plac Gwarków 1, 40-166 Katowice
Since radiation risks are usually considered to be related to nuclear energy, the majority of research on radiation protection has focused on artificial radionuclides in radioactive wastes, spent nuclear fuel or global fallout caused by A-bomb tests and nuclear power plant failures. Far less attention has been paid to the radiation risk caused by exposure to ionizing radiation originating from natural radioactivity enhanced due to human activity, despite the fact that technologically enhanced naturally occurring radioactive materials (TENORM) are common in many branches of the non-nuclear industry. The monitoring and prevention of occupational radiation risk caused by enhanced natural radioactivity TENORM has become obligatory in many cases of industry of concern. Usually the applied radiation risk constrains have been assumed based on primordial rules off radiation protection, exactly the same as in case of risk related to artificial radioactivity. Case-specific risk scenarios have been developed and became available for industry operators. However, less attention has been paid to the environmental impact associated with TENORM residues. Such solid materials are often deposited directly into the environment a practice which is strictly forbidden in the management of other types of radioactive waste. In order to assess environmental impact, information on radionuclide speciation, interaction and transfer within affected ecosystems and spatial and temporal distributions of radionuclide species, influencing their mobility and biological uptake, is often required. In view of existing preliminary datasets, the need to consider TENORM waste as a particular unique case of environmental hazard is quite apparent. In particular, TENORM residues differ significantly in quantity and quality from radioactive materials arising from the nuclear fuel cycle. In addition, the radiation risk is usually combined with the risk caused by other pollutants. This largely explains why TENORM can't be managed directly by applying the rules designed and developed for other categories of radioactive waste. But up to now there are no reasonable and clear regulations in this matter. As a result, the non-nuclear industry is often not aware of potential occupational as well as environmental problems caused by natural radioactivity or they would expect negative consequences in the case of implementing radiation protection measures. The modification of widely comprehended legislation with requirements taken from radiation protection seems to be the first step to solve this problem and raise awareness about enhanced natural radioactivity for all stakeholders of concern.
100 NUTECH-2011
BIOACCUMULATION OF POLONIUM (210Po) AND URANIUM (234U, 238U) IN PLANTS AROUND PHOSPHOGYPSUM WASTE HEAP IN WIŚLINKA (NORTHERN POLAND)
A. Boryło, B. Skwarzec, G.. Olszewski, D. Strumińska-Parulska
University of Gdańsk, Faculty of Chemistry, Analytics and Environmental Radiochemistry Chair, Sobieskiego 18/19, 80-952 Gdańsk
210 234 23 8 In the study the activities of polonium Po and uranium in plants, collected near phosphogypsum waste heap in Wiślinka (northern Poland), were determined by using the alpha spectrometry. The highest amounts of polonium and uranium were found in common wheat (Triticum aestivum) samples. The comparability polonium and uranium contents were confirmed in bedders, but higher accumulation was determined in ripe species than immature species of vegetables. The higher polonium and uranium concentrations were noticed in green part of plant, the lower in root of plants. Polonium concentration in cultivated plants samples was not species diverse. The maximum radionuclides concentrations were found in green part of red beet (Beta vulgaris esculenta), carrot (Daucus carota) and parsley (Petroselinum sativum): 3.50±0.32 Bqkg-1 wet wt for 210Po and 17.27±0.69 Bqkg-1 wet wt for 238U, 2.68+0.11 Bq-kg-1 wet wt for 210Po and 16.29+0.63 Bq-kg-1 wet wt for 238U and 2.11±0.14 Bq-kg-11 wet wt for 21091n Po and 15.60±0.29 Bq-kg-11 O wet wt for U respectively, the minimum in leek (Aliumporrum): 0.83±0.02 Bq-kg-1 wet wt for 210Po and 3.46±0.23 Bq-kg-1 wet wt for 238 210 234 andU .23 8Th U ein result the analyzes obtained dplant for smeado were whighe plantr sin showeroot odf thaplantt ths etha concentrationn in their grees ofn partsPo,. ThUe highest activities 210 23 8 of Po and U were estimated in the ruderal species (fat hen Chenopodium album, gallant-soldier Galinsoga parviflora, common chickweed Stelaria media) (between 87.42+0.93 Bq-kg-1 wet wt and 91.40+0.89 Bq-kg-1 wet wt for 210Po, between 50.05±2.29 Bq-kg-1 wet wt and 78.21±3.27 Bq-kg-1 wet wt for 238U in root, between 49.63±0.63 Bq-kg-1 wet wt and 53.21±0.57 Bq-kg-1 wet wt for 210Po, between 27.89±2.60 Bq-kg-1 wet wt and 42.91+4.00 Bq-kg-1 wet wt for 238U in green parts) as well as willow samples (Salix viminalis) (SRC) from protection zone of phosphogypsum waste heap -1 210 -1 238 (40.98±1.19 Bq-kg wet wt for Po and 67.15±2.21 Bq-kg21 0we t wt fo23r 8 U in green parts of plant). The higher concentrations of natural radionuclides Po and U were estimated for hydrophyte (common sedge Carex nigra Reichard), the favorite habitat of which is particularly wet meadow and for plants collected in the vicinity of phosphogypsum waste heap. The highest uranium and polonium concentrations were characterized for plants, which are covered with tomentose. The tomentose may cover total plant or the some its parts (leaves, stem and fruits). Its task is plant protection against the cold, excessive evaporation or too strong insolation and desiccation by wind. The tomentose strongly scatters sunlight, so some parts of plant are matt and turn silver, grey in color. The main source of polonium and uranium is dry and wet atmospheric fallout in the immediate vicinity of phosphogypsum waste heap and the transfer via root for distant areas. The diverse structures of analyzed plant roots (storage root system, taproot system, fine root system and structural root system) fundamentally influence the amounts of in-taken polonium and uranium forms. Acknowledgments: The authors would like to thank the Polish Ministry of Higher Education and Sciences for the financial support of this work under grant: DS/8460-4-0176-1. 102 Environmental Studies
CURRENT STATUS OF RADON AND RADIUM ACTIVITY MEASUREMENTS IN WATER AT THE FEDERAL UNIVERSITY OF TECHNOLOGY (UTFPR, BRAZIL)
Janine Nicolosi Corrêa, Jaqueline Kappke, Sergei A. Paschuk, Hugo R. Schelin, Valeriy Denyak, Allan F. N. Perna, Marilson Reque
Federal University of Technology - Paraná, UTFPR, Av. Sete de Setembro, 3165, Curitiba, PR, Brazil
Current work describes the present status and obtained results concerning 222Rn activity measurements in drinking water collected from artesian bores at Curitiba region during the period of 2009 - 2010. The measurements were performed at the Laboratory of Applied Nuclear Physics of the Federal University of Technology in cooperation with the Nuclear Technology Development Center (CDTN) of Brazilian Nuclear Energy Committee (CNEN). Experimental setup was based on the Professional Radon Monitor (ALPHA GUARD) connected to specific kit of glass vessels Aqua KIT through the air pump. The equipment was adjusted with air flow of 0.5 L/min. The 222Rn concentration levels were detected and analyzed by the computer every 10 minutes using the software DataEXPERT by GENITRON Instruments. Collected average levels of 222Rn concentration were processed taking into account the volume of water sample and its temperature, atmospheric pressure and the total volume of the air in the vessels. Collected samples of water presented the average 222Rn activity about 60 Bq/L which is almost 6 times more than maximum level of 11.1 Bq/L recommended by the USEPA (United States Environmental Protection Agency). It has to be noted that few artesian drillings presented the radon activity of almost 200 Bq/. Further measurements are planned to be performed at other regions of Parana State and will involve the mineral water sources, explored artesian drillings as well as soil samples. Another subject of this work is the preliminary results concerning 226Ra activity measurements in bottled mineral water offered at the market of Curitiba-PR, Brazil. Experimental setup was based on the Professional Radon Monitor RAD7 (Durridge Company, Inc.). This detector is equipped with special kit of glass vessels and equipment permit to identify the 222Rn activity concentration in small (40mL and 250mL) water samples. Before the measurements, it was verified the background of RAD7 detector together with Radon In Water Accessory (RAD H20) using the samples of distilled water. The evaluation of radium (222266 Ra) soluble salts in water and its activity concentration was performed when 222222R n in water samples was reached the secular equilibrium since its production rate (226Ra activity) became equal to its decay rate (222Rn). For this purpose, collected water samples were stored in hermetic bottles of 250mL during 45 - 50 days before the measurements. The minimum and maximum 226Ra concentration ranges were of 0.03 and 2.95 Bq/L, respectively. Eight samples of bottled mineral water presented values of 226Ra concentration range above 0.1 Bq/L which is the maximum activity level of alpha global radioactivity established by the Norms and Regulation of Brazilian Ministry of Health (Portaria n. 518/2004). Further measurements are planned to perform with other brands of bottled mineral water present at the market of Parana State, Brazil.
Acknowledgments: The authors are very thankful to CNPq, CAPES, CNEN and Fundação Araucária (Paraná St.) for financial support of this work as well as to colleagues from the Institute of Radiation Protection and Dosimetry (IRD) and from the Center of Nuclear Technology Development (CDTN/CNPq) for permanent positive discussions and assistance in the measurements
103 NUTECH-2011
EVALUATION OF THE ORIGIN OF SULFATE IN THE WATER SOURCE KLJUC, SERBIA N. Miljevic1, D. Boreli-Zdravkovic1, G. Dusan2, B. Mayer3 [email protected]
1Jaroslav Cerni Institute for Development of Water Resources, Jaroslava Cernog 80, 11226 Belgrade, Serbia 2Vinca Institute for Nuclear Sciences, POB 522, 11001 Belgrade 3Department of Geocience, University of Calgary, 2500 University Drive NW, Calgary, Alberta T2N 1N4, Canada
34 18 The sulfur and oxygen isotope ratios of sulfate ( SSO4 and OSO4, respectively) in groundwater are commonly used in aquifer studies to identify sulfate sources and describe biogeochemical processes. A sampling campaign was carried out during low flow conditions (September 2007) in central Serbia. Sixteen samples taken from piezometers completed in the aquifer and from the river were analyzed for pH, electrical conductivity (EC), redox potential 2 18 34 18 (Eh), Hwater, Owater, SSO4 and OSO4 as well as major ions. The groundwater pH was neutral (6.8 7.3) at shallow depths (5.7 9.5 m bgl) with redox potential of 339-372 mV. The sulfate concentrations in groundwater samples obtained from the study area varied from 56.2 34 to 165mg/l and SSO4 values ranged from - (Fig. 1).
24 Groundwater V Morava river 20 Sea water Evaporites 16 Atmospheric deposition
an 12 thro po 8 genic
soil 4 5 SO 2- 4 4 0 Sulfate from 3 sulfide oxidation 2 1 -4 sulfate from sulfide oxidation 0
-1 -8 -6 -5 -4 -3 -2 -1 0 1 2 3 4 34S CDT -12 -40 -30 -20 -10 0 10 20 30 40 50 34 S CDT
Figure 1. Schematic diagram of the isotopic composition of typical sulfate sources including data for groundwater and the Velika Morava River in the vicinity of the Kljuc groundwater source.
34 There is a weak trend of increasing SSO4 values with increasing sulfate concentrations in the area indicating that bacterial dissimilatory sulfate reduction is not occurring. A few 34 18 samples with slightly lower SSO4 (< 0 ) have also lower OSO4 (<2 values, suggesting that these may have a component of 34S-depleted sulfate from sulfide oxidation [1]. 1. Hosono T, Chung-Ho Wang, Umezawa Y, Nakano T, Onodera S, Nagata T, Yoshimizu C, Tayasu I, Taniguchi M (2011) Multiple isotope (H, O, N, S and Sr) approach elucidates complex pollution causes in the shallow groundwaters of the Taipei urban area. Journal of Hydrology 397:23-36. 104
Environmental Studies
MEASURMENT OF THE RADON CONCENTRATION OF AIR SAMPLES IN THE SARI CITY
A.Rahimi
Faculty of member, Department of Medical Physics, Mazandaran University of Medical Sciences
A completed radon remediation programme of the type implemented in Northampton shire is most cost-effective for an Action Level between 200 and 300 Bq m 3. The implications for future health policy are discussed. Although the sensitivity of the RTM2100 is lower (1.5 cpm/(kBq/m3) at the expected high humidity levels) and its internal volume is nearly three times higher (370 ml), also this instrument is suitable for water measurements. Both units are working with Alpha spectroscopy and both are equipped with membrane pumps.
Tablel. Density of Radon at Seasons Radon(Bq m-3) Season Number of house 28.615 spring 650 27.20 summer 650 27.07 autumn 650 36.95 winter 650
Table2. Exopsure Dose at Seasons
DOSE(pSv ) Season Number of house 0.032 spring 650 0.026 summer 650 0.037 autumn 650 0.056 winter 650
1. "Protocols for Radon and Radon Decay Product Measurements in Homes." EPA 402-R-92- 003, June, 2009. 2. "Permanent Wood Foundation System - Basic Requirements, NFPA Technical Report No.7." 2010. 3. "Radon Control Options for the Design and Construction of New Low-Rise Residential Buildings," ASTM Standard Guide, E1465-92, 2009. 4. "Radon Handbook for the Building Industry," NAHB-NRC, 2010. 5. "Radon Resistant Construction Techniques for New Residential Construction. Technical Guidance. " EPA/625/2-91/032, February 2010. 6. "Ventilation for Acceptable Indoor Air Quality," ASHRAE 62-2009.
105 NUTECH-2011
MODELLING OF CALENDAR TIMESCALES FOR LAMINATED LAKE SEDIMENTS IN NORTHERN POLAND
N. Piotrowska1, W. Tylmann2, M. Kinder2, D. Enters3
'GADAM Centre of Excellence, Department of Radioisotopes, Institute of Physics, Silesian University of Technology, Gliwice, Poland 2University of Gdansk, Institute of Geography, Gdansk, Poland 3University of Bremen, GEOPOLAR Institute of Geography, Bremen, Germany
In the framework of NORPOLAR (Northern Polish Lake Research; http://www.norpolar.ug.edu.pl/), interdisciplinary research on annually laminated sediments from four northern Polish lakes is carried out. These paleolimnological studies will provide completely new, high resolution data of climatic and environmental changes in this part of Europe. Preliminary studies confirm that sediment records retrieved covers the time span form the Last Glacial (ca.12-16 kaBP), through the entire Holocene up to the present time. The interpretation of research results requires a calendar age assigned to the depths of the sediment, which is called age-depth model. The models for the investigated lakes will be based on different dating methods, taking into account the assumptions, limitations and uncertainties of each method. Varve counting is being performed for laminated sections of the cores. The isotopic methods are also applied. 210Pb dating has a time range ca. 150 years and will provide additional age control for uppermost parts. It will be supported by 137Cs marker. 14 Radiocarbon ( C, T1/2=5730 yrs) dating is used to supplement and verify the varve chronologies, with special regard to poorly-laminated and homogenous parts. The most reliable 14C dates are obtained on terrestrial plant macrofossils deposited in the lake sediments, e.g. fragments of leaves, pine needles, birch seeds. The mass of such samples are relatively small, therefore the Accelerator Mass Spectrometry (AMS) is required for the measurements. This technique, in opposite to radiometric techniques (liquid scintillation and gas proportional counting), allows to determine the 14C/12C ratio in a sample. In the modern biosphere this ratio is around 10-12, and for samples within range of the 14C method it is 103 times smaller. The main advantage of AMS technique is the reduction of the mass of the sample needed to make the measurement from ca. 1 g of carbon to 0.1-1 mg. In order to obtain the 1% uncertainty of measurement the time of several minutes is required, as opposed to several days for the radiometric techniques. This presentation will give an overview of the current state of work and provide the first (preliminary) age-depth models, based on 21 AMS radiocarbon ages, 210Pb ages and partial varve chronologies. The project "Modelling of calendar timescales for laminated lake sediments in Northern Poland as a basis for high-resolution palaeoenvironmental reconstructions" is currently running, with funding for 50 additional 14C samples and 137Cs measurements. Statistical analysis of all available information related to time will be performed in order to construct robust age-depth models and to calculate the corresponding uncertainties.
106 Environmental Studies
NATURAL RADIOACTIVITY IN GROUNDWATER
M. Dulinski, N.D. Chau, P. Jodłowski, J. Nowak, K. Rozanski, M. Sleziak, P. Wachniew
AGH University of Science and Technology, Faculty of Physics and Applied Computer Science, al. Mickiewicza 30, 30-059 Krakow, Poland
Radioactive substances are ubiquitous on Earth. They can be grouped into several distinct classes with respect to their origin: (i) primordial radionuclides, (ii) cosmogenic radionuclides, (iii) radionuclides produced in natural decay series, and (iv) anthropogenic radionuclides. Although environmental aspects of both natural and anthropogenic radioactivity have been widely discussed in scientific literature, presence of natural radioisotopes in groundwater as a hazard factor to the public has been seldom addressed in sufficient detail. Growing use of groundwater resources for drinking water purposes calls for careful evaluation of this aspect of presence of radioactive substances in the environment. This presentation reviews the subject of natural radioactivity in groundwater, with emphasis on those radioisotopes which contribute in a significant way to the overall dose received by the consumers. Literature data on the observed activity ranges of major radionuclides present in groundwater will be presented, followed by discussion of physicochemical factors that control the levels of selected radionuclides in aquifer systems. Radiological aspects of the presence of natural radionuclides in groundwater will be reviewed, including discussion of current regulations dealing with radioactivity in drinking water. Examples of dose calculations received by different age groups in the population as a consequence of consumption of groundwater with specific levels of radioactivity will be presented and discussed.
107 NUTECH-2011
PLUTONIUM SPECIATION IN THE SOUTHERN BALTIC SEA SEDIMENTS
D.I. Strumińska-Parulska, B. Skwarzec, M. Pawlukowska
University of Gdańsk, Faculty of Chemistry, Chair of Analytics and Environmental Radiochemistry, Sobieskiego 18/19, Gdańsk, Poland
The principal source of plutonium radionuclides in the Baltic Sea is the atmospheric fallout from nuclear weapon tests. The other sources: plutonium releases from spent fuel facilities in Sellafield (UK) and Cap de la Hague (France) are less important. Since 26 of April 1986 it's been a new source of plutonium - Chernobyl plutonium, which should be taken under note in estimation of its radiological effects on the environment. In this study 6 different chemical plutonium fractions (dissolved in water, connected to carbonates, connected to oxides, complexed with organic matter, mineral acids soluble and the rest) in sediments from the delta of Vistula River, Gdańsk Bay, Gdańsk Deep and Bornholm Deep were determined. The distribution of 239+240Pu in analyzed sediments samples was not uniform and depended on its chemical form, depth and geomorphology of the Baltic sediments. The highest amount of total plutonium concentration exists in middle parts of sediments, below 5 cm and comes from global fallout from atmospheric nuclear tests in 1958- 61. The highest 239+240Pu concentrations were found in muddy sediments of Gdańsk Bay at 8- 9 cm layer while the lowest, 10 times lower than in Gdańsk Bay, near the delta of Vistula River. Chernobyl-derived plutonium exists in Vistula delta and Gdańsk Bay sediments in 239+240 layer of 3 cm depth. According to all analyzed fractions the biggest amount of pu in all analyzed sediments was connected to fraction connected to carbonates: the delta of Vistula River - 34%, Gdańsk Bay - 42%, Bornholm Deep - 35%. In sediments of Gdańsk Deep plutonium was mostly connected to fraction soluble in mineral acids (40%) and also organic matter (29%).
Acknowledgments: The authors would like to thank the Ministry of Science and Higher Education for the financial support under grant DS/8460-4-0176-1.
108 Environmental Studies
POLONIUM, URANIUM AND PLUTONIUM BIOACCUMULATION IN MARINE BIRDS
D.I. Strumińska-Parulska1, B. Skwarzec1, A. Boryło1, J. Fabisiak2
[email protected] University of Gdańsk, Faculty of Chemistry, Analytics and Radiochemistry Chair, Sobieskiego 18/19. 80-952 Gdańsk, Poland 2Naval Academy, Śniadeckiego 61, 81-103 Gdynia, Poland
The estimation of the size of contamination caused by alpha radioactive elements in the natural environment and their effects on living organisms is one of the most important issues of the radiochemical and radiological protection. Among alpha radionuclides polonium, uranium and plutonium play an important roles because of their strong accumulation in the marine biota and they are an important sources of radiation dose in the body of marine animals. Birds are double-environment animals and they are an important part of the ecosystems. Seabirds are a very important element of the trophic chain of marine ecosystem. Particularly the birds' feathers are often used as a bioindicator of metal and radionuclide contamination of marine and air environment. In our study 11 marine birds species were examined: 3 species of permanently residing at southern Baltic, 4 species of wintering birds and 3 species of migrating birds. The results showed that polonium, uranium and plutonium are non-uniformly distributed in analyzed seabirds. Among all analyzed radionuclides the highest concentrations were noticed in feathers and viscera and the lowest in skin. Further experiments provided, polonium, uranium and plutonium are mostly adsorbed on feathers not built in. About 63% of polonium, 63-67% of uranium and 82% of plutonium are apparently adsorbed suggesting the external sources such as air or water. The bioaccumulation levels of polonium, uranium and plutonium depend not only on their concentrations in the food but also their concentration in the natural environment. Seabirds are typical double-environment (land-aquatic) animals and radionuclides can come from water and air. That is why seabirds feathers can be good radiological biomonitoring indicator and the isotopic composition of adsorbed radionuclides reflects environment pollution.
Acknowledgments: The authors would like to thank the Polish Ministry of Higher Education and Sciences for the financial support of this work under grant: DS/8460-4-0176-1.
109 NUTECH-2011
RADIATION DEGRADATION OF OIL POLLUTED SOILS AND ITS COMPONENTS
D. Abbasova
dinara_abasova@hotmail. com
Institute of Radiation Problems of AzNAS, Azerbaijan
Oil and coal extraction, transports pollution, fires, intensive development of various branches of industry, all of the above factors exert a negative effect on the environment and therefore, on humans. This environment impact presents a hazard to human health in cancer, allergic and other diseases manner. That is why pollutants degradation is one of the important ways to decrease its contents in environment. Polycyclic Aromatic Hydrocarbons (PAHs) are belong to persistent organic pollutant and have very carcinogen and toxic properties. For PAHs degradation was selected irradiation technology. In this work gamma irradiation for PAHs degradation in oil-polluted soils was used. In contaminated soil, PAHs concentration changes during irradiation from 200 to 500 kGy for almost all PAHs are small and irregular (that is G ~ 0). This is not surprising, because in a heterogeneous system exist a "protective effect" owing to the lower efficiency of energy transfer to possible aggregation of impurities and the influence of microstructure (such effects in other systems are well known, they even offered to use, for example, to protect the active ingredients in Radiation sterilization of drugs). Thus, we conclude that the radiation-chemical conversion of most of the studied PAHs in experimental conditions is negligible. The exceptions are naphthalene, anthracene and phenanthrene. For naphthalene at doses up to 400 kGy observed expenditure (decreased concentration) starting material with the slowdown, and with further exposure - slight changes. Perhaps, naphthalene is partially converted into more condensed aromatic hydrocarbons by an ionic mechanism, or reacts with OH radicals and hydrogen atoms formed during irradiation of the other (non-aromatic) components of the mixture. It is also possible partial oxidation, which was confirmed by infrared spectra. For anthracene and phenanthrene is observed, conversely, increase in concentration at doses up to 400 kGy, and then a marked decline at 500 kGy. Growth can be attributed to the formation of more stable anthracene and phenanthrene from a "light" products, in particular, from the same naphthalene. In conclusion, we thing that irradiation can be used like degradation method, but doze should be higher than 500 kGy.
110 Environmental Studies
RADIOCARBON FOR NUCLEAR ENERGY
A. Pazdur, N. Piotrowska, K. Tudyka
Centre of Excellence - Gliwice Absolute Dating Methods Centre, Institute of Physics, Silesian University of Technology, Gliwice, Poland
The 14C isotope was identified as a product of nuclear reactions between thermal neutrons and nitrogen nuclei (reaction 14N(n,p)14C) at the University of California, Berkeley in the 1930s [1], Activity of 14C in modern biosphere is close to 226 Bq (kgC)"1. Radiocarbon decays to 14N through the ß" decay, with the maximum energy of 156 keV and the half-life equal to 5720±30 years. In the nuclear reactor the 14C is produced in the reactions of thermal and fast neutrons with N, O and C nuclei, which are present in the surrounding materials, and as a product of nuclear fuel decay. Since the 1980's the monitoring of 14C content in the vicinity of nuclear power plants and nuclear reprocessing facilities has revealed the elevated rise in 14C for a number of sites. The average production rate of 14C in the atmosphere in the vicinity of nuclear reactor was estimated by [2,3] to ca. 3.40 TBq GW"1 yr"1, while ca. 0.94 TBq is released to the atmosphere. The numerous examples of 14C content monitoring in the atmosphere and biosphere can be found [2,4,5,6]. The studies prove that the area of elevated 14C content extends to ca. 10-20 km, and the maxima were found around few hundred meters form the plant chimney stack. The considerable rise of 14C content was also observed during the Chernobyl accident, where the 14C activity was ca. 900 Bq kg"1 [7], The measurements of 14C content changes in the surrounding of nuclear facility may be used as a tool for monitoring of the neutron component of nuclear radiation in the case of normal reactor work, as well as during its malfunction. The systematic measurements of 14C content in the atmosphere and organic material like short-lived plants, tree leaves, annual tree rings, snail shells, enable determination of its superfluous production on the time and space scales.
1. Libby, WF (1967) History of Radiocarbon Dating. In: Radioactive Dating and Methods of Low-level Counting, proceedings of a symposium held in Monaco, 2-10 Mar 1967, International Atomic Energy Agency, Vienna (Austria), 744 pp. 2. Otlet RL, Walker AJ, Fulker MJ (1990) Survey of the dispersion of 14C in the vicinity of the UK reprocessing site at Sellafield. Radiocarbon 32:23-30 3. Otlet RL, Walker AJ, Longley H (1983) The use of 14C in natural materials to establish the average gaseous dispersion patterns of releases from nuclear installations. Radiocarbon 25:593-602 14 4. Levin I, Munnich KO, Weiss W (1980) The effect of anthropogenic CO2 and C sources on the distribution of 14C in the atmosphere. Radiocarbon 28:634-643 5. Segel M, Levin I, Schoch-Fischer H, Munnich M, Kromer B, Tschiersch J, Munnich KO. (1983) Anthropogenic 14C variations. Radiocarbon 25:583-592 6. Keogh SM, McGee EJ, Gallagher D, Mitchell PI (2004) Spatial and temporal impacts of 14C releases from the Sellafield nuclear complex on the Irish coastline. Radiocarbon 46;2:885-892 7. Kovaliukh NN, Skripkin VV, Awsiuk R, Pazdur A, Pazdur MF, Los IP, Buzinny MG, Nesvetailo VD (1994) Zapis emisji radiowęgla w czasie awarii reaktora jądrowego w Czarnobylu w słojach rocznych przyrostów drzew. Zesz. Nauk. Politechniki Śląskiej seriaMat.-Fiz. 71, Geochronometria 10:217-224
111 NUTECH-2011
RADIOCHEMICAL ANALYSIS OF RADIUM 226Ra IN ENVIRONMENTAL SAMPLES USING ALPHA SPECTROMETRY
A. Boryło, B. Skwarzec, G. Olszewski, D.I. Strumińska-Parulska [email protected]
University of Gdańsk, Faculty of Chemistry, Analytics and Environmental Radiochemistry Chair, Sobieskiego 18/19, 80-952 Gdańsk, Poland
Presented radioanalytical procedure concerns radium analysis in environmental samples from the area of phosphogypsum waste in Wiślinka, Poland. Phosphogypsum samples were used to develop this methodology as this residue from the process of producing phosphoric fertilizers from phosphorites is known to contain significant amounts of natural radionuclides including 226 224 radium isotopes 226Ra and 224Ra. This property makes it a good material for modifying the determination of radium. Radiochemical analysis of radium isotopes in environmental samples contains four steps: digestion, ion exchange chromatography, radium electrolysis and finally alpha spectrometry measurements. Phosphogypsum samples were digested in the mixture of concentrated HNO3 and HCl. Radium was separated on three exchange columns: Dowex 1x8 100-200 mesh anion exchange resin in 9M HCl (uranium washed with 60 ml of 0.5M HCl), Dowex 1x8 50-100 mesh anion exchange resin in 8M HNO3 (thorium and plutonium washed with 60 ml of 8M HCl) and cationic resin Dowex 50Wx8 50-100 mesh in 2.5M HCl (radium separation from barium). Radium fraction was stripped with 125 ml of 6M HCl and heated to dryness. The residue was dissolved in 0.17M ammonium oxalate at pH 2.6 adjusted with few drops of concentrated HN03 and radium was electrolyzed on stainless steel disc with 400 |ig of platinum as potassium hexachloroplatinate(IV). The electrolysis was 90 min with current intensity kept constant 700 mA. Activities of 226Ra in analyzed samples were measured using alpha spectrometer Alpha Analyst S470, Canberra-Packard.
Acknowledgments: The authors would like to thank the Ministry of Science and Higher Education for the financial support under grant DS/8460-4-0176-1.
112 Environmental Studies
RADIOISOTOPE INVESTIGATIONS OF COMPOUND TWO PHASE FLOWS IN OPEN CHANNEL
M. Zych1, L. Petryka2, J. Kępiński3, R. Hanus4, T. Bujak5, M. Śleziak2, E. Puskarczyk1
^GH - University Science and Technology, The Faculty of Geology, Geophysics and Environmental Protection, Department of Geophysics 2AGH - University Science and Technology, The Faculty of Physics and Applied Computer Science, Department of Applied Nuclear Physics 3AGH - University Science and Technology, The Faculty of Geology, Geophysics and Environmental Protection, Department of Universal Geology, Environmental Protection and Geotouristic 4Rzeszow University of Technology, The Faculty of Electrical and Computer Engineering, Department of Metrology and Diagnostic Systems 5Rzeszow University of Technology, The Faculty of Mathematics and Applied Physics, Department of Physics
The paper presents advantage of radioisopes application to investigation of dispersed small solid particles transportation in open channels. In such flow continuous sedimentation removes significant part of solid phase from the stream. Superposition of those processes significantly complicating the flow often observed in Carpathian rivers. Such flows are modeling in the laboratory stand (fig. 1) constructed in Sedimentological Laboratory of The Faculty of Geology, Geophysics and Environmental Protection AGH. Ultrasonic, and radioisotope methods supplemented by advanced signal analysis [3,4] are introduced to recording of solids velocity and concentration distribution in the flow gives verification of the proposed models.
/ I - open channel 2 - feeding container 3 - spillway container 4 - pump 5 - feeding pipeline 6 - sluice-valves 7 - throttle valve 8 - conveyer of granular substance 9 • universal can of measurement 10 - ultrasound flow meter 11 - watermark Figure 1. Scheme of the installation at the AGH - University Science and Technology
1. Petryka L, Zych M, Murzyn R (2005) The non-stationary two-phase flow evaluation by radioisotopes. Nukleonika 50:43-46 2. Zych M, Petryka L, Hanus R (2010) Radioisotope evaluation of two phase liquid-solid flow in a vertical pipe. Measurement Automation and Monitoring PAK 56:315-321 3. Petryka L, Oszajec J (1993) The cross-correlation method of solid state particle velocity measurements in industry. Nuclear Geophysics 7;2:323-333 4. Petryka L, Hanus R, Zych M (2008) Statistical signal analysis in the radioisotope two- phase flow measurements. Measurement Automation and Monitoring PAK 54:863-868
113 NUTECH-2011
SOURCE APORTIONMENT OF PARTICULATE MATTER (PM10) COLLECTED IN KRAKOW, POLAND
L. Samek, G. Fira
Faculty of Physics and Applied Computer Science, AGH University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow, Poland
Samples of PM10 were collected in winter 2010 at two different sites in Krakow, Poland. One was located nearby a steel factory (in the Nowa Huta district) and the other was situated at a distance of 10 km from the first one (in the Krowodrza district). The detected mass concentrations varied between 33 and 320 (j,g/m3. The daily limit value is equal to 50 (j,g/m3. The Environmental Protection Agency was in charge of both - collecting the samples and providing the mass concentrations. Measurements of elemental concentrations and statistical analyses were performed at the University of Science and Technology. Mean concentrations of Ti, Cu and Br were almost the same for both sites. Mean concentrations for K, Ca, Cr, Zn and Pb were two times higher for the site in Nowa Nuta district than they were for the one in Krowodrza district. For Mn three times and for Fe four times higher values of mean concentrations were evaluated for the Nowa Huta district. Source contributions to ambient PM10 were determined by factor analysis (FA) and multilinear regression analysis (MLRA) based on PM10 composition data which included elemental concentrations. During winter time the main contributors to pollution of PM10 have been local combustion, industry and other non-identified sources such as secondary aerosols. For the Nowa Huta district 53.1% of sources were identified as combustion, 28.5% as industry and 18.3 non-identified. For the Krowodrza district, industry contributed 50.4%, combustion 46.1% and non-identified 3.5% to pollution of PM10. These results are presented in the figure 1. While looking at the meteorological parameters we saw that concentration of K was inversely correlated to temperature for both sites. Potassium could be coming from wood combustion. Similar percentage contributions of pollution sources have been observed at both sites. Yet higher concentration of various elements has a negative influence on the industrial localization.
3,5
28,5
1 1
50,4 • 2 2 46,1 3 3
53,1
a b Figure 1. Source contributions at a) Krowodrza and b) Nowa Huta 1 - Industry, 2 - Combustion, 3 - Non identified
114 Environmental Studies
SPATIAL DISTRIBUTION OF EQUIVALENT GAMMA DOSE RATE IN THE VICINITY OF MINE WATER SEDIMENTATION PONDS, UPPER SILESIAN COAL BASIN
M. Śleziak, M. Duliński
monika.sleziak@gmail. com
AGH - University of Science and Technology, Faculty of Physics and Applied Computer Science, al. Mickiewicza 30, 30-059 Kraków
Highly mineralized waters from the Upper Silesian Coal Basin Region mines after pumping on the surface are stored in sedimentation ponds. They contain elevated concentrations of natural radionuclides, mainly radium, which is partly removed from the water column due to precipitation and co-sedimentation and is stored in the bottom sediments. Fluctuations of water level in the sedimentation ponds lead to exposure of sediments to the atmosphere. Consequently, sediment particles containing radium can be transported due to wind action and deposited around the ponds thus increasing gamma dose rate above the natural background. To estimate the radiological hazard associated with presence of sedimentation ponds, three such objects have been selected for further investigations. Two of them, Brzeszcze and Kaniów are in operation and receive mine waters from Brzeszcze and Silesia mine, respectively. The third pond (Rontok Duży) has been in use between 1977 and 1997 and was collecting waters from Silesia Mine. Relatively large repository of gangue adjoins the Kaniów pond. The results of equivalent gamma dose rate measurements cover relatively wide range of values. Over the Kaniów pond embankment the measured dose rate was between 0.13 and 0.35 [j,Sv/h (3.1 mSv/year). Directly over the sediments, the measured values were between 0.59 and 8.37 [j,Sv/h (73.4 mSv/year). In the Rontok Duży pond the gamma dose rate measured over the embankment was between 0.07 and 0.22 [j,Sv/h (1.9 mSv/year), and directly over the sediments - between 0.47 and 1.09 [j,Sv/h (9.6 mSv/year). The dose rates in the vicinity of the Rontok Duży pond were significantly lower than in the vicinity of Kaniów pond. The lowest dose rates were observed around the Brzeszcze pond. Gamma dose rates in the embankment were between 0.07 and 0.17 [j,Sv/h (1.5 mSv/year). Over the gangue repository the dose rates are in the range from 0.13 to 0.33 [j,Sv/h (2.9 mSv/year). In uncontaminated area the measured dose rates did not exceed 0.11 [j,Sv/h. The mean annual dose rate received general public in Poland is equal 3.5 mSv/year. The highest measured gamma dose rates in the vicinity of the studied ponds were more than 20 time higher than this reference value.
115 NUTECH-2011
TRACE ANALYSIS OF VOLCANIC ASH AND IT'S LEACHING DYNAMICS
S. Landsberger1, B. Canion1 and C. Jacques
1 University of Texas at Austin, Nuclear Engineering Teaching Lab, R-9000, Austin, Texas, USA 78712 2 École Nationale Supérieure d'Ingénieurs de Caen, Caen France
There is continued great interest in determining the trace element and heavy metal content of volcanic ash for a variety of reasons. The motivation stems from the understanding the geochemistry of volcanic ash in imbedded geological formations, the impact on seawater, and the possible release of toxic elements into the environment such livestock grazing and weathering indices. We have employed Compton suppression neutron activation analysis with thermal and epithermal neutrons to determine trace elements in volcanic ash from Indonesia's Mount Merapi eruption in October 2010. In addition we performed leaching dynamics with the US Environmental Protection Agency Toxicity Characterization Leaching Procedure (TCLP), simulated rainwater and seawater to ascertain the potential leaching dynamics of the trace elements and heavy metals. These methods and techniques have given a broader range of elements that can be normally determined by routine neutron activation analysis and a better understanding of the impact of these heavy metals have on the environment.
116 Environmental Studies
URANIUM AND RADIUM ISOTOPES IN GEOTHERMAL WATERS
Nguyen Dinh Chau, J. Nowak
AGH University of Science and Technology, al. Mickiewicza 30, 30-059 Kraków, Poland
Based on the collected data from the our measurements and the attained references, the correlation between temperature, mineralization and concentrations of the uranium and radium isotopes in the thermal waters were made and are presented on the suitable Figs 1,2 and 3. The figures show that there is neither dependent of mineralization on temperature and nor of uranium, radium isotopes concentrations on temperature. The fact can be explained by differences between the leaching processes of the rock minerals into waters and the temperature transmission of the heat energy from the rock formation. General one can state that, in the most of geothermal groundwaters (90%) the 226Ra activity contents are larger than that of 228 Ra ones (Fig. 4), so the 226 Ra is easier to be leached than 228 Ra under high temperature. But the radium content increases with mineralization, though every relation characterizes the features of the geological and hydrological conditions of the region, where the question groudwater occurs.
80 • Espain O France * Tunesia • Poland o France • Poland 60
ra W £ ca £ 40 QW
20 a • ° cn ff 0 # i» • 0 20 40 60 80 100 0 0 20 40 60 80 100 o Temperature [ C] Temperature [oC] Figure 1. Relation between temp. and TDS Figure 2. Relation between temp. and 238U
2500
2500 O France * Tunesia • Poland 2000
m 1500 O.SQT 1500 E, m[ * * a Cl 1000 , * * 0£ 1000 r • • « O • 500 y** 0 0 500 1000 1500 2000 2500 0 0 20 40 60 80 100 2288 r Temperature [oC] Ra [mBq/L]
Figure 3. Relation between temp. and 226Ra Figure 4. Relation between 226226Rt a and 228228RT a
117 NUTECH-2011
ACTIVATION OF LINEAR MEDICAL ACCELERATORS - AN OVERVIEW
K. Polaczek-Grelik1, B. Karaczyn2
1Medical Physics Department, University of Silesia, Poland department of Nuclear Physics and Its Applications, University of Silesia, Poland
Electromagnetic radiation of the energy above 10 MeV effectively interacts with atomic nuclei causing photonuclear reactions. Due to rapid increase in cross sections of (y,p), (y,n), ) interactions in various materials, particle flux as well as radioactive nuclei are generated. In this paper depletion-activation model [1] is applied for qualitative analysis of originating radionuclides and their contribution to the evolution in time of the overall level of linac activity including periods when the therapeutic beam is off or is emitted. Taking into consideration the dynamics of the radionuclides generation during emissions of a therapeutic beams and their decay in-between, the activity of a linac during 24 h-period has been described. This characterization includes linacs with three different photon modes (15, 18 and 20 MV) working in radiotherapeutic facilities in Poland. Semiconductor gamma-ray spectrometry was used as a suitable method for identification of individual radionuclides that make up a complex radioactive source - a head of the high- energy linear accelerator. The method of analysis allows for estimation of the flux of photons with defined energy and hence, to quantify the dose at the chosen point in the vicinity of linac, using the flux-to-dose conversion coefficients [2]. Doses stated this way were checked with the indications of a standard dose rate meter for consistency. Since the cross sections of nuclear reactions are strong energy-dependent [3], the apparent activity increase with the nominal potential of linac, what is presented in Table 1. Near the head casing the apparent ambient dose equivalent is 1.5-2 times lower than under the head - in front of the collimator system. Careful choice of the sequence of low- and high-energy beam emission as well as short- and long-time procedure could lower the activity cumulated during the working day.
Table 1. Ambient dose equivalent H*(10) [jaSv/h] connected with radionuclides originating inside an accelerator head estimated immediately after the end of 2620 MU of X-ray emission and 10 minutes after. The major contribution of nuclides originated via neutron capture (n,y) reactions is evident.
Linac type Primus Clinac 2100 Clinac 2300 Nominal potential 15 MV 18 MV 20 MV immediately 17.4 ±4.3 21.8 ± 7.8 87.6 ± 14.3 main source :8Al (79.0 %) 6Mn (40.3 %) 56Mn (61.0 %) Near the head casing after 10 min 18.5 ±6.7 60.1 ±9.8 main source 2Br (41.6 %) 6Mn (45.3 %) 6Mn (85.0 %) immediately 28.0 ±7.0 46.0 ± 17.5 150.4 ±20.4 main source 28Al (84.9 %) 187W (31.2 %) 187W (28.8 %) Under the head after 10 min 43.9 ± 16.7 147.3 ± 19.9 main source 6Mn (24.9 %) 187W (32.5 %) 187W (29.3 %)
1. Abdel-Rahman W, Podgorsak E (2005) Neutron-activation revisited: The depletion and depletion- activation models. Medical Physics 32;2:326-336. 2. ICRP Publication 74 (1996) Conversion coefficient for use in radiological protection against external radiation. Pergamon Press, Oxford, 159, 175. 3. ENDF/B-VII.0 data base (2006). 120 Radiotherapy
APPLICATION OF THE MONTE CARLO CALCULATIONS FOR A DESIGNING OF THE SIMPLE ENERGY MODULATOR IN THE PASSIVE BEAM-DELIVERY TECHNIQUE FOR THE 50 MEV - 70 MeV PROTON BEAMS
M. Grządziel, A. Konefał, W. Zipper
Institute of Physics, Department of Nuclear Physics and Its Application, University of Silesia, Katowice, Poland
Protons exhibit little scattering when they penetrate matter and give the highest energy deposition (called the Bragg peak) near the end of their range [1]. However, energy of protons has to be modulated to get the uniform dose distribution in the whole tumor volume. The modulation of the proton energy makes it possible to get the spread-out Bragg peak and to obtain the maximum dose in the range of the appropriate depths. In general, there are two beam-delivery techniques to get the spread-out Bragg peaks: passive and dynamic [2]. In the passive technique the proton beam is scattered and degraded in a set of absorbers. In the dynamic beam-delivery technique the beam is moved by a magnetic field across the tumor cross-section whereas a change of the depth of penetration is achieved by a change of energy of protons. The purpose of the presented investigations was the use of the Monte Carlo calculations (GEANT4 code [3,4]) for a designing of the simple proton energy modulator in the passive beam-delivery technique. The Monte Carlo method is not used in the treatment planning systems, because of the insufficient calculation power of the contemporary computers. However, the Monte Carlo calculations deliver very sensitive results. The proton beams with energies ranging from 50 MeV to 70 MeV with energy spread of 2 % were considered in the presented investigations. Such beams are used for treatment of the well-localized surface tumors. The designed system for the passive beam-delivered technique is a structure consisting with 2 leaves modulating energy of protons. The leaves are made of 3 PMMA ([-CH2-C(CH3)(COOCH3)-]n, 1.18 g/cm ) and differ with their thickness. The thickest leaf has the thicker of 1.68 cm whereas the thinner one is 3.36 mm thick. This energy modulator is a structure geometrically similar to stairs created by the leaves. The depth-dose distributions were calculated in a cubic water phantom. The proton beam hits the center of the phantom surface. The appropriate motion of the energy modulator (i.e. a shift of the leaves in relate to the beam) makes it possible to achieve the spread-out Bragg peak. To simplify the calculations and to reduce the time of computer simulations, the primary needle-like proton beams (i.e. the beams with no spatial spread) and no collimators were considered. The designed modulator of the proton energy makes it possible to obtained the spread-out Bragg peaks with the width of 4 mm, independently on the proton energy. The ratio of the minimum dose to maximum dose is less than 0.4.
1. Konefał A, Szaflik P, Zipper W (2010) Influence of the energy spectrum and the spatial spread of the proton beams used in the eye tumor treatment on the depth-dose characteristics. Nukleonika, 55(3): 313-316. 2. Journal of the ICRU, Report 78, Vol. 7 No 2 (2007), Oxford University Press. 3. http://geant4.web.cern.ch 4. Konefał A (2006) Monte Carlo simulations with the use of the GEANT4 code. Postępy Fizyki 57; 6:242-251, (in Polish). 121 NUTECH-2011
AUTOMATION SYSTEM FOR QUALITY CONTROL IN THE MANUFACTURING OF IODINE-125 SEALED SOURCES USED IN BRACHYTHERAPY
Samir L. Somessari, Anselmo Feher, Francisco E. Sprenger, Maria Elisa C. M. Rostellato, João A. Moura, Osvaldo L. Costa, Wilson A. Parejo Calvo
[email protected]; [email protected]; [email protected]; [email protected]; [email protected]; [email protected]; [email protected]
National Nuclear Energy Commission Institute for Nuclear and Energy Research IPEN-CNEN/SP Av. Prof. Lineu Prestes, 2242 05508-000 - Sao Paulo, SP - Brasil
The aim of this work is to develop an automation system for Quality Control (QC) in the production of iodine-125 sealed sources, after undergoing the process of Laser Beam Welding (LBW). These sources, also known as iodine-125 seeds are used, successfully, in the treatment of cancer by brachytherapy, with low-dose rates. Each small seed is composed of a welded titanium capsule with 0.8 mm diameter and 4.5 mm in length, containing iodine-125 adsorbed on an internal silver wire. The seeds are implanted in the human prostate to irradiate the tumor and treat the cancerous cells. The technology to automate the quality control system in the manufacturing of iodine-125 seeds consists in developing and associate mechanical parts, electronic components and pneumatic circuits to control machines and processes. The automation technology for iodine-125 seed production developed in this work employs Programmable Logic Controller (PLC), step motors, drivers of control, electrical-electronic interfaces, photoelectric sensors, interfaces of communication and software development. Industrial automation plays an important role in the production of Iodine-125 seeds, with higher productivity and high standard of quality, facilitating the implementation and operation of processes with Good Manufacturing Practices (GMP). Nowadays, the Radiation Technology Center at IPEN-CNEN/SP imports 36,000 iodine-125 seeds per year and distributes them for clinics and hospitals in the country. However, the Brazilian potential market is of 8,000 iodine-125 seeds per month. Therefore, the local production of these radioactive seeds has become a priority for the Institute, aiming to reduce the price and increase the supply to the population in Brazil.
122 Radiotherapy
DESIGN AND PERFORMANCE OF A SYSTEM FOR TWO- DIMENSIONAL READOUT OF GAS ELECTRON MULTIPLIER DETECTORS FOR PROTON RANGE RADIOGRAPHY
W. Dąbrowski, T. Fiutowski, B. Mindur, P. Wiącek, A. Zielińska
Faculty of Physics and Applied Computer Science, AGH University of Science and Technology, Krakow, Poland
The Proton Range Radiography (PRR) technique is expected to provide significant improvements in accuracy of the hadron therapy, which is presently limited by the uncertainty in stopping power distribution. The technique requires measuring residual energies and positions of mono-energetic protons passing through the target. Such an imaging system can be operated in-situ before and after the treatment allowing real time monitoring of the irradiated tissue position. A detector system suitable for such applications must be capable of measuring the residual proton energy after passing the object to be imaged and sub-millimetre spatial resolution over the area of 30 x 30 cm2. In order to limit the exposure time and obtain real-time images the detector and data acquisition must provide capability of working with high-count rates up to 106/s. Several detection technologies are investigated to realize a system meeting these requirements. The PRR detector system being developed comprises two Gas Electron Multiplier (GEM) detectors to determine trajectories of protons and a stack of 30 thin plastic scintillators to measure the energy losses [1]. In this paper we present a novel concept of two-dimensional readout of the GEM detectors, which provide high count rate capability as required for PRR application being developed in collaboration with the TERA Foundation. The key component of the system is an Application Specific Integrated Circuit (ASIC) named GEMROC. The GEMROC comprises 32 independent channels, each one capable of measuring the amplitude and the time of incoming signals. The signal amplitudes and corresponding time stamps measured with 4 ns resolution are stored in the derandomizing buffers. The buffers are read out via a token-based multiplexer, which provides on-chip zero suppression. Reconstruction of the hit positions is performed in an external FPGA based data acquisition module by matching the time stamps of signals recorded in X- and Y-strips. The amplitude information is used for finding centers of gravity for clusters of signals on neighboring strips belonging to the same detection events. Such a readout architecture utilizing data derandomization and zero suppression can cope with the required count rates up to 106/s compared to 104/s achieved in the existing prototype instrument. Performance of the GEMROC ASIC has been evaluated by electronic measurements as well as in the test bench comprising a GEM detector. The presented test results confirm that the developed GEMROC ASIC and data acquisition system fulfill the requirements for readout of the GEM detectors in the PRR instrument.
1. Amaldi U, et al. (2011) Construction, test and operation of a proton range radiography system. Nucl. Instr. andMeth A 629: 337-344.
123 NUTECH-2011
DETERMINATION OF ENERGY SPECTRA OF THERAPEUTIC X-RAY BEAMS FROM MEDICAL LINACS FOR VARIOUS IRRADIATION CONDITIONS
M. Bakoniak, A. Konefal
Institute of Physics, University of Silesia, Department of Nuclear Physics and Its Application, Katowice, Poland
The energy spectra in water phantom for the therapeutic high-energy X-ray beams are not easy to determine because the experimental methods are very difficult to perform whereas the Monte Carlo calculations need the suitable computer power. This fact is evidenced by the lack of extensive data including the energy spectra in water. Even dosimetry protocols do not include such kind of information. Authors present usually X-ray spectra determined in air. However, the spectra in air cannot be generally used to characterize accurately the beam quality in another irradiated medium. In our investigations the energy spectra for the 6MV X- ray beam were determined along the beam central-axis in water - a medium recommended by the dosimetry protocols. The spectra were derived using the Monte Carlo method for open and wedged fields. The GEANT4 code with the LowEnergy models of interactions of photons with matter was applied for the calculations. All calculations were performed under the linux operation system, on computers in the Department of Nuclear Physics and Its Application of Institute of Physics of University of Silesia in Katowice (Poland). The simulation program was verified by the comparison of the calculated depth-dose characteristics in water with those measured with the use of the Markus ionization chamber. The spectra were calculated in the range of the depths up to 27.35cm for SSD=100cm. This work showed that the shapes of the spectra as well as the mean energy Ed of the beam depended strongly on the radiation field size and depth d in water phantom. The detailed knowledge of the energy spectra of therapeutic beams from medical linacs is essential for the calculations of the stopping power ratios or the beam quality correction factors and for dose calculation algorithms in advanced treatment planning systems, for investigations of treatment machine head design etc.
124 Radiotherapy
FETAL DOSE EVALUATION IN BREAST RADIOTHERAPY USING SHIELDING AND PHYSICAL AND ENHANCED DYNAMIC WEDGES
Filipov, D.12, Schelin, H. R.1,2, Soboll, D.S.1
danifilipov@gmail. com
1 Federal University of Technology-Paraná, UTFPR, Curitiba-PR, Brazil 2 Pelé Pequeno Príncipe Research Institute, IPP, Curitiba-PR, Brazil
When a pregnant woman is submitted to breast radiotherapy, the fetus may be seriously affected by the peripheral dose [1]. In order to verify that dose, a humanoid phantom, was irradiated at the left breast. The phantom is an adapted manikin, with some materials (densities close to water) inside and outside of it, as shown in figure 1.
O nŁR <=S>
Figure 1. A) nylon rod; B) the rod being placed in one of the holes; C) all the rods (internally drilled) placed in the holes; D) the humanoid phantom The irradiation was done using a 6 MeV x-ray beam energy from a linear accelerator "Clinac 600C". During the irradiation, a shield around the abdominal area of the manikin, consisting of blocks and slabs of lead was used. In addition, two types of filters were used: a physical, with 30o angulation, and an enhanced dynamic one. Figure 2 shows the irradiation layout.
Figure 2. A) enhanced dynamic wedge, B) physical wedge being used during the phantom irradiation Through a cylindrical ionization chamber, positioned in the simulator's fetal region (inside of the nylon rods), it was found that, at the end of the breast treatment, the peripheral doses reach values between 3.90 and 48.67 cGy, when the physical wedge was used. With the application of the enhanced dynamic wedge, the values were between 1.75 and 13.78 cGy. According to the obtained data, the physical wedge can increase the peripheral dose due to the larger background radiation intensity and to the scattering caused by the attenuator material. In addition, the shielding couldn't block all the secondary radiation, which, according to the literature, can be able to induce mental retardation and cancer during postnatal life. However, the induction to these effects is negligible, when the type of wedge was changed. [2]
1. Filipov, D.; Mafra, K.C.; Schelin, H.R.; Soboll, D.S. Fetal Dose Evaluation in X-Ray Radiotherapy in Cases of Advanced Gestation. WC 2009, IFMBE Proceedings 25/III, pp. 519- 522. 2. Stovall, M.; et al. Fetal Dose From Radiotherapy With Photon Beams: Report Of AAPM. Radiation Therapy Committee Task Group No. 36. Medical Physics, v.22, p.63-82, 1995. 125 NUTECH-2011
IMPROVEMENTS IN THE QUALITY CONTROL OF IRIDIUM-192 WIRE USED IN BRACHYTHERAPY
Osvaldo L. Costa1, Carlos A. Zeituni1'2, Maria Elisa C. M. Rostelato1, João A. Moura1, Anselmo Feher1, Eduardo S. Moura1, Carla D. Souza1, Samir L. Somessari1
1 Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN - SP Avenida Professor Lmeu Prestes, 2242 ZIP 05508-000 São Paulo, SP - Brazil. 2 Instituto Presbiteriano Mackenzie Rua da Consolação, 930 ZIP 01302-907 São Paulo, SP - Brazil
Brachytherapy is a method used in the treatment of cancerous tumors by ionizing radiation produced by sources introduced into the tumor area, this method seeks a more direct attack to the tumor, thereby maximizing the radiation dose to diseased tissue while minimizing the dose to healthy tissues (1). One of the radionuclides used in brachytherapy is iridium-192. The Radiation Technology Center (CTR) of the Nuclear and Energy Research Institute (IPEN) has produced commercially, since 1998, iridium-192 wires used in low dose rate (LDR) brachytherapy (2). To produce this radionuclide, firstly a iridium-platinum wire is irradiated in the nuclear reactor IEA-R1 for 30 hours with a neutron flux of 5 x 1013 ncm-2s-1, the wire is left to decay by 30 days to remove the main contaminants and then goes through a quality control before being sent to the hospital. In this quality control is checked the radiation homogeneity along each centimeter of the wire (3). To implement this procedure is used a device consisting of an ionization chamber surrounded by a lead shield with a small 1 cm wide slit, linked to the ionization chamber is a voltage source and a Keithley 617 electrometer, 2 minutes is the range used to measure the charge by the electrometer. The iridium wire is considered in accordance when there is no variation greater than 5% between the average measures and the maximum and minimum values. However, due to design features of the measurement system, the wire may appear to the detector through the slit in larger sizes than the ideal, improperly influencing the final quality control. This paper calculates the difference in size of these variations in profile and their influence on the final count, it compares the actual values obtained and describes the improvements made in quality control procedures that provided more accurate measurement data, analyzes the results and suggests changes in devices aimed at further improving the quality control of iridium-192 wires produced at IPEN and used in hospitals in Brazil.
1. Air d EG, Williams JR, Rembowska A (2000) "Brachytherapy", Radiotherapy Physics in Practice (Williams, JR, Thwaites, DI, Eds). Oxford Univ. Press, Oxford 2. Rostelato, MECM (1997) Preparação de fontes de iridio-192 para uso em braquiterapia. Dissertation (MSc.). Instituto de Pesquisas Energéticas e Nucleares, São Paulo 3. Rostelato MECM, Rela PR, Zeitune CA, Feher A, Manzoli JE, Moura JA, Moura ES, Silva CPG (2008) Development and production of radioactive sources used for cancer treatment in Brazil. Nukleonika; 53(Supplement 2):S99-S103
126 Radiotherapy
MONOLITHIC APPLICATORS OF 125I AND 106Ru APPLIED IN EYE CANCER BRACHYTHERAPY
I. Cieszykowska, A. Piasecki, T. Janiak, M. Żółtowska ,T. Barcikowski, M. Mielcarski
Institute of Atomic Energy Radioisotope Centre POL ATOM, 05-400 Otwock/Świerk
AIM. Ophthalmic applicators therapy is widely used treatment of intraocular tumors. Presented paper shows method for preparation of seed-less 125I and 106Ru ophthalmic applicators according to the own, innovative design. The background of the idea was the preparation of uniformly adsorbed 106Ru or 125I on a concave surface of silver shell, which is afterwards hermetically sealed inside the acrylic material. Such an insert can be mounted in typical metallic capsules used in 125I seed holding applicators. METHODS. For the deposition of radionuclides on silver, a convenient method of internal electrolysis was applied. For 125I a cell containing silver anode and platinum cathode was chosen. Electrolyte solution contained the mixture of 0.01-0.1 M NaOH, varying concentration of NaI added as a carrier followed by carrier- free 125I. Solution of 106Ru and ruthenium nitrosyl-trichloride in sulphamic acid was electrolyzed in a cell with silver cathode and aluminium anode. The cathode was separated from the anode by a separator made of polyethylene foil doped with silica. The ruthenium carrier concentration in the catholyte was 0.5-^2.0 g/L in 4% wt. sulphamic acid, whereas the anolyte was a solution of 4% wt. sulphamic acid. The radioactive silver shell was hermetically sealed between two layers of acrylic material. Active inserts having the same shape and dimensions as the seed-holding insert, after passing contamination tests according to ISO 9978, were mounted in the same metallic capsules as used with the seed applicators. RESULTS. Test showed a higher quality of 125I deposit obtained using electrolyte without addition of Nal as a carrier. Deposition yield of 125I reached 97-98% after the process lasting 70-90 h. The depth dose rate measurements indicated that the total radioactivity incorporated in a seed-less applicator is about 3 times lower than that in seed-containing one, while simultaneously assuring the expected dose rate. The measurement of 125I surface distribution uniformity on a silver shell, related to the value in the centre of applicator which is assumed to be 100%, indicated that deviation from uniformity was 6.8 %. The highest efficiency of 106Ru deposition achieving 80% after electrolysis lasting 24 h was obtained by the use catholyte containing 106Ru and 0.7 mg/mL ruthenium as carrier. The deposits obtained were metallic and lustrous, adhering well to the underlying silver. The activities of prepared 106Ru applicators sealed in acrylic material were between 13 and 20 MBq depending on type, which assured the required surface dose rate of about 120 mGy/min. Based on tests performed in accordance with ISO 2919, ophthalmic applicators with monolithic active core were classified as C 33222. CONCLUSION. Quality of seed-less applicators of 125I and 106Ru with monolithic active core produced by Radioisotope Centre of IEA POLATOM was confirmed by CE marking. The primary advantages of offered devices in comparison with seed-containing applicators are as follows: uniformity of radioactivity distribution in active core, contributing to improvement of therapeutic properties of the source, decreased radioactivity of the source maintaining the required dose rate, possibility of using the same types capsules for active cores containing various radionuclides (106Ru, 125I)
127 NUTECH-2011
A NEW MONTE CARLO SUPPORT FOR THE INTERPRETATION OF THE GAMMA-GAMMA BOREHOLE GEOPHYSICAL TOOL RESPONSES IN CASE OF HETEROGENEOUS BOREHOLE VICINITY
U. Wiącek1, T. Żorski2, U.Woźnicka1, D.Dworak1
:The Henryk Niewodniczański Institute of Nuclear Physics Polish Academy of Sciences, Kraków, Poland 2 AGH University of Science and Technology, Faculty of Geology, Geophysics and Environmental Protection, Kraków, Poland
A spectrometric gamma-gamma density logging is a basic borehole method in geophysical prospecting. A typical tool consists of Cs-137 gamma source and two ("near" and "far") detectors. An essential method for the interpretation of logs is a "spine & ribs" chart. This methodology was primarily introduced, in the sixties of the last century, to correct the mud cake influence on responses of the gamma-gamma compensated (two detectors) density tools. However, other challenges arrive when density tool is implemented for measurements in deviated or horizontal holes, either as wireline log or LWD (Logging-While-Drilling). In these cases the radial symmetry, common in vertical holes, is lost. In the presented paper we try to explore a little bit more this, above mentioned, pure geometrical effect. A signal of the gamma-gamma density tool in specific borehole conditions has been numerically calculated. Transport of gamma rays, from a point Cs-137 gamma source situated in a borehole tool, through rock media to the tool detectors, has been simulated using a Monte-Carlo code. An influence of geometrical heterogeneity in the rock medium surrounding the borehole on the signal of the detectors has been examined. The authors focused on a simple borehole-rock model, where axial symmetry is disturbed by the presence of two rock layers of different density and porosity. A flat boundary between these two layers, parallel to the borehole axis, is assumed. As a result of calculations, a few full range ribs were plotted in the "spine & ribs" chart, giving very instructive pictures. The ribs were found on both sides of the spine. Distances between the borehole wall and the consecutive two-layer boundaries were changed from 0 to 20 cm - this scope ensures exhibition of the rib bows in their full ranges. An additional rib was calculated for full radial symmetry case (where two cylindrical layers differ only in densities), obtaining similar result as for the case of the flat boundary. Effect of changing the absolute density values (for the bordered layers) on the total tool responses has been also analyzed. As a conclusion one can say: on the basis of responses of the gamma-gamma density tool, using the classic interpretation procedure "spine & ribs", it is possible to assess the distance between the border separating media of different densities, and the borehole wall. The presented examples of calculations once more confirm wide possibilities of the Monte Carlo computer simulations for solving nuclear well logging problems, especially in case of complex borehole geometry.
130 Radiometric Measurements
ANALYSIS OF NATURALLY OCCURING RADIOACITVE MATERIAL USING NEUTRON ACTIVATION ANALAYSIS AND PASSIVE COMPTON SUPPRESSION GAMMA-RAY SPECTROMENTRY
S. Landsberger1, G. George2, D.Tamalis3, J. Jean-Louis3
1 University of Texas at Austin, Nuclear Engineering Teaching Lab, R-9000, Austin, Texas, USA 78712 2Enviroklean Products Development, Inc., Houston, Texas, USA 77064 3Florida Memorial University, Department of Health and Natural Sciences, Miami Gardens, Florida USA 33054
Naturally occurring radioactive material (NORM) still remains a problem in oil and gas exploration. Radioactive wastes from oil and gas drilling take the form of produced water, drilling mud, sludge, slimes, or evaporation ponds and pits. In many parts of the USA the soils contain radioactivity that are then concentrated in mineral scales on the pipes, storage tanks and other extraction equipment. The radionuclides Ra-226 and Ra-228 are the primary radionuclides in the waste. We are investigating the use of neutron activation analysis (NAA) to determine if any other heavy metals may also concentrate in the NORM wastes as well as using passive radioactivity counting using Compton suppressed gamma-ray spectrometry. With a low-energy germanium counter and the Compton suppression system much lower detection limits can be achieved to measure Ra-226, Ra-228 and their daughter products. A full explanation will be given highlighting the Compton system as well as the results from NAA.
131 NUTECH-2011
APPLICATION OF SEALED RADIOACTIVE SOURCES TO MEASURE AND CONTROL FLOW IN INDUSTRIAL PROCESSES
L. Petryka1, H.J. Pant2, M. Zych3, D. Kosman1, M. Śleziak1, R. Hanus4
1AGH University of Science and Technology, Department of Applied Nuclear Physics, Krakow, Poland 2Isotope Applications Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085, India 3AGH University of Science and Technology, Department of Geophysics, Krakow, Poland 4Rzeszow University of Technology, Department of Metrology and Diagnostic Systems, Rzeszow, Poland
Knowledge of flow rate is an essential requirement in chemical, petroleum and petrochemical industries as well as in many harbor installations. The information about flow rate is needed for a number of purposes, such as calibration of installed flow meters, assessment of efficiency of pumps/turbines and estimate of material balance etc. Conventional methods such as ultrasonic are often used for continuous and online measurement of flow rates and estimate of material balance in industry. However, it is not practically feasible to apply these methods in a situation where different chemicals are pumped through a single pipeline one after other and where the pipeline is partially filled with liquid during flow. Presently, a cross-correlation method based on sealed radioactive sources in conjunction with conventional ultrasonic gauge is proposed. Signals provided by both the gauges facilitate determination of varying flow rates and total volume during pumping or transfer operation. The method is noninvasive and can find applications in various industrial situations. The schematic diagram of the technique is shown in Fig. 1. The paper describes details of the developed technique and its implementation.
Correłometer
i FLOW t" d -X
1 - sealed source 2 - scintillation probe 3 — ultrasonic flowmeter
Figure 1. Schematic diagram of the proposed flow rate measurement in a pipeline
1. Petryka L, Hanus R, Zych M, Sleziak M (2010) Two phase flow measurement by radioisotopes. Electrical Review 86;5:24-29, 2. Zych M, Petryka L, Hanus R (2010) Radioisotope evaluation of two phase liquid-solid flow in a vertical pipe. Measurement Automation and Monitoring PAK 56:315-321, 3. Petryka L, Hanus R, Zych M (2008) Statistical signal analysis in the radioisotope two- phase flow measurements. Measurement Automation and Monitoring PAK 54:866-868.
132 Radiometric Measurements
DEVELOPMENT OF THE MECHANICAL SYSTEM ON A THIRD- GENERATION INDUSTRIAL COMPUTED TOMOGRAPHY SCANNER IN BRAZIL
Wilson A. Parejo Calvo, Carlos H. de Mesquita, Francisco E. Sprenger, Fabio E. da Costa, Pablo A. Vasques Salvador, Diego V. de Souza Carvalho, Margarida M. Hamada
[email protected]; [email protected]; [email protected]; [email protected]; [email protected]; [email protected]; [email protected]
National Nuclear Energy Commission Institute for Nuclear and Energy Research IPEN-CNEN/SP Av. Prof. Lineu Prestes, 2242 05508-000 - Sao Paulo, SP - Brasil
The development of measurement geometry for medical X-ray computed tomography (CT) scanners carried out from the first to the fourth-generation. This concept has also been applied for imaging of industrial processes such as pipe flows or for improving design, operation, optimization and troubleshooting. Nowadays, gamma CT permits to visualize failure equipment points in three-dimensional analysis and in sections of chemical and petrochemical industries. The aim of this work is the development of the mechanical system on a third- generation industrial CT scanner to analyze laboratory gas absorption column which perform highly efficient separation, turning the 60Co, 137Cs or 192Ir sealed gamma-ray source and the NaI(Tl) multidetector array. It has also a translation movement along the column axis to obtain as many slices of the process flow as needed. The mechanical assembly for this third- generation industrial CT scanner is comprised by strength and rigidity structural frame in stainless and carbon steels, rotating table, source shield and collimator with pneumatic exposure system, spur gear system, translator, rotary stage, drives, and stepper motors. The use of suitable spur gears has given a good repeatability and high accuracy in the degree of veracity. The data acquisition boards, mechanical control interfaces, software for movement control and image reconstruction were specially development. This third-generation industrial CT scanner has obtained good spatial resolution and images. The filtered back projection (FBP) tomographic reconstruction algorithm used has shown a faster convergence. The mechanical system presented a good performance in terms of strength, rigidity, accuracy and repeatability with great potential to be used for education or program dedicated to training chemical and petrochemical industry professionals and for industrial process optimization in Brazil.
133 NUTECH-2011
EXPERIMENTALLY AND NUMERICALLY PREDICTED RESIDENCE TIME DISTRIBUTION IN CHEMICAL REACTORS
Z. Stęgowski1, L. Furman1, C.P.K. Dagadu2
1 Faculty of Physics and Applied Computer Science, AGH University of Science and Technology, Krakow, Poland 2National Nuclear Research Institute, Ghana Atomic Energy Commission, Accra, Ghana
The quantitative description of mixing is a key subject of chemical reaction engineering. The concept of residence time distribution (RTD) arising from tracer experiments can be used to quantify mixing. The experimental RTD allows determining some global parameters as mean residence time, standard deviation and flow detection through stagnate zones or bypasses. More fundamental information concerning flow patterns and mixing can be obtained through numerical simulation tools - computational fluid dynamics (CFD) for example. Increasing numbers of studies on flows in reactors using CFD have been reported in the last decade. Currently, CFD methods constitute very powerful tools and are in common use for flow patterns prediction. Anyway, in majority of cases the CFD results have to be validated by experimental measurements. The experimental RTD method, among others, can be used for this validation. In CFD codes the RTD function can be simulated utilizing the particle tracking method. Since the RTD concept is not common in fluid dynamics studies, many CFD codes require some supported software for RTD simulation. The paper will present the particle tracking method for RTD simulation using FLUENT® CFD software supported by MATLAB® codes. Two case studies comparing the experimental and simulation RTD results will also be presented. The first will present the laboratory jet mixer and the second - the industrial gold leaching tank study.
Experimental and simulated tracer flow visualization. Progress of dye tracer in the reactor at 8, 16 and 24 s after pulse feed.
134 Radiometric Measurements
IMAGING TECHNIQUE FOR TROUBLESHOOTING OF INDUSTRIAL EQUIPMENT BY GAMMA-RAY ABSORPTION SCANS
Marcio I. Haraguchi1, Hae Y. Kim2, Francisco E. Sprenger3, Wilson A. Parejo Calvo3
[email protected]; [email protected]; [email protected]; [email protected]
lrrricom Technology Ltda., Av. Conselheiro Rodrigues Alves, 58 - Centro, 12.620-000 - Piquete, SP - Brazil 2Sao Paulo University (U SP) Polytechnic School - Department of Electronic Systems Engineering (LPS) Av. Prof. Luciano Gualberto, Tr. 3, 158, 05508-900 - Sao Paulo, SP - Brazil 3National Nuclear Energy CommissionInstitute for Nuclear and Energy Research IPEN-CNEN/SP Av. Prof. Lineu Prestes, 2242, 05508-000 - Sao Paulo, SP - Brasil
Gamma scanning is one of the most common nuclear techniques on troubleshooting industrial equipments like distillation columns and reactors. With a very simple concept, the technique is easy to implement. Searching for a competitive edge the industry has been long developing solutions to achieve better results. On the last decades, significant development has been done with the advent of new equipments, electronics, portable computers and software, to the point that nowadays the field work and reporting can be done in a question of hours. Continuous scanning and wireless detection systems are examples of successful field solutions, while new software aid on reporting and data presentation. However the type and quality of the results itself has not dramatically changed since its beginning. A scan profile is simple to understand, although the process to build it can be very complex as it requires a specific blend of knowledge and abilities. Process Engineering, Chemical Engineering, Internal Hydraulic Project, Nuclear Engineering and field abilities are pre requisites for of any scan specialist rookie. Correct data gathering, interpretation and reporting are abilities often difficult to match or requires a long time of training. Probably there are no more than a handful of scan specialists on the world. The industry faces a similar difficult on the customer side, as it is always necessary to train end users to understand a report and how to use its best. This paper describes our effort on developing a new approach on the gamma scan test using image reconstruction techniques that would result on a graphic image rather than a XY plot. Direct and easier to understand, a report with graphic images would be also accessible to a wider audience, not limited to the customers experienced with gamma scan interpretation.
135 NUTECH-2011
LABORATORY AUTOMATIC MEASURING SYSTEM OF GAMMA SPECIMENS
P. Filipiak, A. Jakowiuk, J. Bartak, B. Machaj, P. Pieńkos, E. Kowalska
Institute of Nuclear Chemistry and Technology, 03-195 Warsaw, ul. Dorodna 16
Gamma counter is designed for low activity measurement of radioactive isotope iodine 125I in liquid or solid samples during radioimmunoassay (RIA) and immunoradiometric assay (IRMA) in small or medium size clinic laboratories. Apart from RIA and IRMA programmed procedures, a measurement of small activity gamma samples is possible. Well scintillator NaI(Tl) is used as radiation detector for isotope 125I in the measuring channel, well dimensions cf> 17x38 mm. Gamma counter LG was adapted for measuring radioactivity of single samples. The counter is equipped with an automatic system allowing transportation of examined samples to scintillator well. Gamma Counter LG is based on an integrated computer which is working under the control of Windows CE and is equipped with special software allowing for setting measurement parameters and communication with an external computer. The measured sample is placed inside scintillator well either by automatic sample feeder or manually. The pulses from photomultiplier tube after amplification and shaping in pulse amplifiers are counted in the window of a single channel analyser by s programmable pulse counter under the control of microprocessor system. The measured count rate is the measure of an activity of the measured sample. Expected count rate error due to unstable operation of amplification is lower than 1%.
136 Radiometric Measurements
RADIOTRACERS AS AN EFFECTIVE TOOL FOR MEMBRANE PROCESSES INVESTIGATION
A. Miskiewicz, G. Zakrzewska-Trznadel
Institute of Nuclear Chemistry and Technology, Warsaw, Poland
There are many methods used for investigation of phenomena taking place in membrane permeation but very often they have fundamental restrictions, e.g. optical methods require membrane modules used in experiments to be transparent or having windows made of glass. Whereas other techniques like Small-Angle Neutron Scattering (SANS), Magnetic Resonance Imaging (MRI) or Electrical Impedance Spectroscopy (EIS) [1-3] are expensive and requiring sophisticated equipment. Radiotracer techniques are the alternative for the study of processes proceeded inside the membrane apparatus. They have a number of advantages; primarily they do not require special or sophisticated equipment to be used. Furthermore, radiolabelled compounds have an advantage over non-active tracers because of very high sensitivity of detection, which gives the opportunity for using very low concentration of the tracer, as well as for remote detection of radiation through the layers of the materials present in the apparatus. In the paper the possibilities of application of radiotracers for determination of effectiveness of the separation process, the rate of the decontamination and accumulation of separated particles on the membrane surface and inside the membrane matrix to examine membrane fouling, were described. The experiments concerned characterization of the cake-layer formed on the membrane: its thickness and the rate of its formation in situ, during the filtration process. Two radionuclides: Ga-68 and La-140 were used as the tracers of the process suspension, which was bentonite suspended in water. Flat-sheet membrane made of PES, with pore size of 0,1 |im was used in this study. The first step of the investigation was determination of conditions of preparation of radiolablled suspension, as well as examination of the stability of Ga-68 and La-140 sorption on the bentonite. Afterwards, the influence of such process parameters like pressure, linear velocity and feed concentration on formation of the bentonite layer on the membrane surface was studied. The results of experiments showed strong reduction of deposit thickness with decrease of concentration of the suspension. The influence of liquid linear velocity in the range under investigation was negative and less significant, however the influence of transmembrane pressure on the deposit thickness was not observed. The method under investigation has proved as a tool for the study of kinetics of the particle deposition and layer formation on the membrane surface. Supportive tools like radiotracers techniques are very helpful in scientific studies, as well as in common, industrial practice. Acknowledgement: The studies were supported by the POIG project No 01.01.02-14-094-09-00 "Analysis of the possibility of uranium supply from indigenous resources" and by the National Center for Research and Development (NCBiR) Research Grant No. R05-058 06/2009. 1. Hamachi M, Mietton-Peuchot M (1999) Experimental investigation of cake characteristics in crossflow microfiltration, Chem. Eng. Sei. 54; 4023-4030. 2. Ensminger D, (1988) Ultrasonics, Marcel Dekker, New York. 3. Chilcott TC, Chan M, Gaedt L, Nantawisarakul T, Fane AG, Coster HGL (2002) Electrical impedance spectroscopy characterization of conducting membranes I. Theory, J. Membrane Sci. 195; 153-167
137 NUTECH-2011
SPECTRAL ANALYZES OF LIQUID-GAS MIXTURE FLOW IN PIPES
L. Petryka1, M. Zych2, A. Sokulska1, R. Hanus3
1AGH University of Science and Technology, Department of Applied Nuclear Physics, Krakow, Poland 2AGH University of Science and Technology, Department of Geophysics, Krakow, Poland 3Rzeszow University of Technology, Department of Metrology and Diagnostic Systems, Rzeszow, Poland
Predominantly multiphase flows in pipes originating many difficulties in analyzes and control. For example, transportation of the same gas volume by a liquid may be performing in form of isolated great bubbles or in form of numerous small one. Moreover, in the last case, transportation of bubbles may take any form from separated clusters to even distribution of bubbles in the pipe. Each of these patterns provoke the various energy losses and interactions between phases, what may be important for a process. So far, these kinds of transport can be assessed visually and only in transparent pipes. The paper shows how gamma radiation can be applied to the analysis of the stream and how it can provide discrete signals. Advance processing of these signals allows determination of gas phase contribution in the flow. In addition, the spectrum of those signals reveals harmonics representing both sizes and arrangement of bubbles in the pipe. Detail investigation of these phenomena were possible due to construction a special rig presented in Fig. 1. The paper brings up results describing in detail the proposed method and its verification by pictures of the transported bubbles. Due to that the method may be easy disseminated in the concerned laboratories.
water
Figure 1. Schematic diagram of the experimental installation
1. Petryka L, Hanus R, Zych M, Sleziak M (2010) Two phase flow measurement by radioisotopes. Electrical Review 86;5:24-29, 2. Petryka L, Zych M, Murzyn R (2005) The non-stationary two-phase flow evaluation by radioisotopes. Nukleonika 50:43-46, 3. Petryka L, Hanus R, Zych M (2008) Statistical signal analysis in the radioisotope two- phase flow measurements. Measurement Automation and Monitoring PAK 54:863-868.
138 Radiometric Measurements
ULTRA-LOW ENERGY X-RAY CALIBRATION SOURCE WITH THE X-RAY TUBE
P. Mazerewicz, W. Czarnacki, A. Gój ska, M. Kisieliński, M. Słapa, M. Traczyk
The Andrzej Solían Institute for Nuclear Studies, 05-400 Otwock-Świerk, Poland
The isotopic sources of y or X-ray with energies below 5 keV, to conduct the calibration process, with the acceptable times of life, virtually non-exist. This absence can be filled by the X-ray source with special X-ray tubes. We have two possible solutions to the calibration sources using X-ray tubes: correction filter method and secondary target method. The Institute for Nuclear Studies developed the low energy X-ray tube, a control unit of X-ray tube and complete head project. The first transmission X-ray tubes with Ag target and Be window dedicated to study of electron beam optics and control characteristics of the X-ray tube were performed. The simulations of the X-ray spectra of the X-ray tube were carried out for different anode voltage and different configuration (target - window - correction filter). The measurements of the X-ray spectra of the above mentioned X-ray tubes for the two solutions of calibration sources were carried out. In order to achieve the desired calibration lines of the source the analysis of the optimal design of the X-ray tube and electron beam parameters were carried out. In the method of correction filters we have a purely apparatus source, which the energy peak is determined by the anode voltage and correction filters, while a stream of photons is determined by anode current of the X-ray tube. In the secondary target method energy peak is determined by discrete fluorescent line of secondary target material and the photon flux intensity is determined by voltage and anode current of the X-ray tube. In this method we can get the sources with two or more calibration peaks. The novel type calibration X-ray source can provide a calibration peaks in the energy range 1.5 keV to 9 keV. This source was already used to calibrate the GEM detectors. Moreover, it can be used to calibration of Gafchromic dosimetric films and TLD detectors as well as to study the radiobiological effects of the ultralow X-rays.
139 NUTECH-2011
188W/188Re GEL GENERATOR
Marcin Konior, Edward Iller
Instytut Energii Atomowej POLATOM, Ośrodek Radioizotopów, Otwock-Świerk, Poland
Rhenium-188 belongs to the group beta-gamma emitters. Radiometric characteristics of radiation emitted by rhenium-188; half life (16.9h ), high energy beta radiation (Ep max=2.118 MeV), low abundance (15.8%) of 155 keV protons created advantageous conditions for medical applications of this radionuclide. The radiopharmaceuticals and radiochemicals containing rhenium-188 are used in clinical trials such as cancer radioimmunotherapy, palliation of skeletal bone pain, endovascular brachytherapy as well as in the pre -clinical development of novel radiopharmaceuticals Its high energy beta particles is sufficient to irradiate medium or large tumors, while the low energy and abundances gamma photons are suitable for imaging. A new approach to preparation of packing of chromatographic column of tungsten- 188/rhenium-188 generators is application of nanocomposites obtained by mean of the sol-gel technique. A specific CSGP method for synthesis of these materials has been elaborated at INCT Warsaw Poland. The initial stage of the process is preparation of the ascorbate- NH4+ - tungsten, next separately prepared zirconyl are added gradually to the reaction mixture. After gelation step, the gels are thermally treatment. This way the synthesis of nanocomposites containing of ZrO2-WO3 with different proportions of oxides were carried out.. For determination of structure of Zr02 - W03 composites the neutron scattering and X-ray diffraction analysis have been applied. The zirconium-tungstate gel in which the oxide molar ratio was 1:2 and annealed at 300oC has been selected for further investigations. The sample of gel with mass 2.9g has been irradiated in Maria nuclear reactor in thermal neutron flux 1.7*1014 cm"2 s"1 for 400 hours. After irradiation the target allowed to "cool" for three weeks. The gamma spectrometry measurements of gel samples performed Ge(Li) detector coupled with multichannel analyzer showed following radionuclides Zr-95, Nb-95m, Re-188, W-188, Hf-188, Co-60, Re-186. The cool gel was packed into chromatographic column of generator which has been washed up with 30 ml of saline and 20 ml of M HCl. Next elution of Re-188 with 10 ml of saline was performed. To evaluate of Re-188 elution profile the samples of 1 ml of eluates were collected and counted in ionization chamber. Using the gamma-ray spectrometry radioisotope composition of eluate has been determined . For the purpose of purification of eluate, the generator was connected to small alumina column. For increasing of Re-188 specific activity, a two system of columns have been tested, first consisted of silver column and anionite column with SepPak QMA, second consisted of silver column and anionite column with Plus QMA. Using these system 85.3% and 88.9% increase of Re-188 activity was obtained. At the present, in scope of project N R05 0062 06 "Żelowy generator izotopowy W-188/Rel88" the experiments with zirconium -tungstate gel composites syntheses by using of WO3 enriched with W-186 are carried out.
142 Radiopharmaceuticals and Radioisotope Production
DEVELOPMENT OF 177LU-PHYTATE COMPLEX FOR RADIOSYNOVECTOMY
Hassan Yousefnia, Amir R. Jalilian, Samaneh Zolghadri, Ali Bahrami-Samani, Mohammad Mazidi1, Mohammad Ghannadi Maragheh
Radiopharmaceutical Research and Development Lab (RRDL), Nuclear Science and Technology Research Institute (NSTRI), Tehran, Iran, Postal code: 14155-1339
Various radiolabeled monoclonal antibodies have been developed for the treatment and diagnosis of malignancies [1-3],Many b—emitters such as 153Sm, I66H0, etc., can be produced in reasonable amounts using (n, gamma) reactions.Due to the half-life limitations in the application of mentioned radionuclides the emerging need for a long halflife beta emitter such as 177Lu is obvious.In this work Lu-177 of 2.6-3 GBq/mg specific activity was obtained 13 2 by irradiation of natural Lu203 sample with thermal neutron flux of 4 x 10 n.cm" .s~'. The product was converted into chloride form which was further used for labeling of 177Lu-phytate complex successfully with high radiochemical purity (>99.9 %, ITLC, MeOH: H2O: acetic acid, 4: 4: 2, as mobile phase). The complex stability and viscosity were checked in the final solution up to 7 days. The prepared complex solution (100 |iCi/100 |il) was injected intra- articularly to male rat knee joint. Leakage of radioactivity from injection site and its distribution in organs were investigated up to 7 days. Approximately, all injected dose has remained in injection site 7 days after injection. The complex was proved to be a feasible agent for cavital radiotherapy in oncology and rheumatology.
1. Zhi Y, Meiying Z, Baohe L, Yan H, Aping M, Qing Z, Xiaobao X (1996) Direct labelling of anti-gastric cancer monoclonal antibody 3H11 with 99mTc. J. Radioanal. Nucl. Chem. 206: 59-67. 2. Ramos Suzarte M, Rodri 'guez N, Oliva JP, Iznaga N, Perera A, Morales A, Gonzalez N, Torres O, Rodri 'guez T (1999) A murine monoclonal antibody for diagnostic of epithelial tumors.J. Radioanal. Nucl. Chem. 240:499-503. 3. Liu N, Jin J, Zhang S, Mo S, Yang Y, Wang J, Zhou M (2001) 211At labeling of a monoclonal antibody and its Fab fragment Cytotoxicity on human gastric cancer cells and biodistribution in nude mice with tumor xenografts. J. Radioanal. Nucl. Chem. 247:129-133.
143 NUTECH-2011
IRIDIUM-192 SEED DEVELOPMENT FOR OPHTHALMIC CANCER TREATMENT
M.E.C.M.Rostelato1, F.R. Mattos1, C.A.Zeituni1, C.D.Souza1, J.A.Moura1, E.S.Moura1, A.Feher1, O.L.Costa1, F. S. Peleias Jr 1, J.R.O. Marques1, R. Belfort Neto 2
Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN - SP Professor Lineu Prestes Avenue, 2242, 05508-zil. 2 Escola Paulista de Medicina, Av. Sena Me 04021-001 São Paulo, SP - Brazil.
Considered a public health problem in Brazil, cancer is the second leading cause of mortality by disease, representing 13.2% of all deaths in the country [1]. Ophthalmic brachytherapy involves inserting an acrylic plate with radioactive material in the eyes of a patient for treatment of ocular tumors. This work is a partnership between Escola Paulista de Medicina - UNIFESP and the Instituto de Pesquisas Energéticas e Nucleares - IPEN for development and implementation of a cheaper therapeutic treatment for ophthalmic cancer with a iridium-192 source, to attend a greater number of patients. Iridium-192 is produced in nuclear reactor. It has a half-life of 74.2 days and decays by beta emission with average energy of 370 keV. [2,3]. The seed will be a platinum-iridium alloy core (80/20), encapsulated in a titanium tube [4]. This project will be divided into the following steps: characterization of materials by FRX (X- ray fluorescence) e EDS (Energy Dispersive Spectroscopy); iridium irradiation in the nuclear reactor IEA-R1; sealing of iridium-192 seed; leakage tests of iridium-192 source in accordance with standard ISO-9978 (radiation protection- Sealed radioactive sources- Leakage test methods) [5]; metallographic tests and measure the activity of the source. The evaluation for use in the ophthalmic treatment of cancer will be made later.
1 BRASIL. MINISTERIO DA SA ÚDE. INSTITUTO NACIONAL DO CÁNCER. Estimativa de incidencia de cáncer no Brasil 2005. Rio de Janeiro; 2005. 2 Oliveira VC, Soares WE; Salvajoli JV, Peres O, Morales FC, Fujisawa FMA (1992) Iridium, terapia versátil, táticas e técnicas. Radiol. Bras., v. 15, n.l, p.44-48 3 Norman S (1965) Iridium-192 as a Radium Substitute. Am.J.Roentgenol. Radium Ther. 93: 170-178 4 Rostelato MECM, Rela PR, Zetuini CA, Feher A, Manzoli JE, Moura JA, Moura ES, Silva COG (2008) Develepment and production of radioactive sources used for cancer treatment in Brazil. Nukleonika 53: 99-103 5 INTERNA TIONAL STANDARD ORGANIZA TION. Radiation protection - Sealed radioactive sources - leakage test methods. Feb. 15, 1992. (ISO 9978).
144 Radiopharmaceuticals and Radioisotope Production
NUCLEAR DATA FOR THE CYCLOTRON PRODUCTION OF IRON-55 VIA VARIOUS REACTIONS
M. Sadeghi1, N. Soheibi2, T. Kakavand2
1Agricultural,Medical and Industrial Research School, Nuclear Science and Technology research Institute,P. O. Box31485/498, Karaj, Iran 2University of Zanjan, Physics Faculty, Zanjan, Iran,
55 , The radionuclide Fe (7 1/2=2.73a) decays by electron capture and consists of small percentage of weak gamma rays [1].This radionuclide can be employed for industrial, medical and agriculture applications. Excitation function of iron-55 via 55Mn(p,n)55Fe, 55Mn(d,2n)55Fe and 54Fe(a,n2p)55Fe reactions were calculated and investigated by TALYS1.0 [2] and ALICE/ASH [3] codes. The required thickness of the target was calculated by SRIM code and theoretical yield of 55Fe production reactions were obtained by means of SIMPSON numerical integral method. The experimental and calculated data were compared with each other. Suitable reaction, 55Mn(p,n)55Fe, was suggested to take full benefite of excitation function and to avoid formation of radioactive and non-radioactive impurities as far as possible.Acording to Fig.1, optimom energy range were predicted to be at 1 MeV to 18 MeV and theoretical production yield were obtained to be 0.3552 (MBq/ |iAh) . To prepare the target, Manganese dioxide (MnO2) powder were used. A thick layer was deposited on an elliptical copper substrate by means of sedimentation method [4] .
• TALYS1.0 A Abyade et al (2010) • ALICE-ASH 800
700 m • • 60 0 • f • ^ 500 ft D o A •• ti 400 •_ o 8H A A. • i/£> 300 AA » O . Q9 Ò 200 A 9 8 o ^ ^mAH 99991 g g g y
0 5 10 15 20 25 Proton energy (MeV) Figure. 1: Excitation function of 55Fe(p,n)55Mn reaction calculated by TALYS 1.0 and ALICE/ASH codes and experimental data
1. Al- Abyad M, Spahn I, Qaim S.M (2010) Experimental studies and nuclear model calculations on proton induced reactions on manganese up to 45 MeV with reference to production of 55Fe, 54Mn and 51C. Applied Radiation and Isotopes 68; 2393-2397 2. Koning A .J et al (2007) TALYS1.0 Nucl. Res . Consul. Group 3. Konobeyev A. Y, Korovin Y. A, Pereslavtsev P. E (1997) Code ALICE/ASH for calculation of excitation functions, energy and angular distributions of emitted particles in nuclear reactions.obninsk institute of Nuclear Power Engineering 4. M. Sadeghi, Z. Alipoor, T. Kakavand, (2010) Target preparation of RbCl on a copper substrate by sedimentation method for the cyclotron production of no-carrier-added 85Sr for endotherapy, NUKLEONIKA,55;
145 NUTECH-2011
NUCLEAR MODEL CALCULATION ON CHARGE PARTICLE INDUCED REACTION ON TI TARGET AND TARGETRY FOR 48V PRODUCTION
M. Sadeghi1, T. Kakavand 2, Z. Ansari2
Agricultural & Medical & Industrial Research Karaj, ^University of Zanjan, Physics Faculty,Zanjan, Iran
48 V( t1/2 =15.98 d ) decay via positron (49.5%) and electron capture(50.4%).It can be used for transmission scanning PET[1]. In this Study 48V excitation function for nat/49/48Ti(p,x)48V and 48Ti(d,2n)48V nuclear reactions were calculated by TALYS 1.0 and ALICE/ACH codes. Recommended thickness of the targets according to SRIM-2006 code was calculated and the theoretical integral yields were computed for all reactions by means of computer software. 48Ti(p,n)48V reaction was determined as the best reaction. Enrich titanium target was used to produce 48V throughout accelerator proton bombardment. Ti target was prepared by sedimentation method [2]. According to SEM scans, enrich titanium target of high-quality morphology prepared using sedimentation method. Observation of neither crack formation nor pit on the surface indicated a good adhesion for the cyclotron purposes.
ALICE/ACH -B-Levkoskij -at- Talys - l.O
soo ŒI c i Ö 400 m A A
1 1 1 1 15 20 25 Proton energy (MeV)
48 48 Excitation function of Ti(p,n) V reaction Calculated by TALYS 1.0, ALICE/ASH codes and experimental data
1. R.D Hickwa., Vanadium-48: A renewable source for transmission scanning with PET, Nuclear physics, 99, 804-806, 1995. 2. M. Sadeghi, Z. Alipoor, T. Kakavand, (2010) Target preparation of RbCl on a copper substrate by sedimentation method for the cyclotron production of no-carrier-added 85Sr for endotherapy, NUKLEONIKA,55;
146 Radiopharmaceuticals and Radioisotope Production
PREPARATION AND QUALITY CONTROL OF 166Ho-DTPA- ANTICD20 FOR RADIOIMMUNOTHERAPY Samaneh Zolghadri, Amir R. Jalilian, Hassan Yousefnia, Ali Bahrami-Samani, Simindokht Shirvani-Arani, Mohammad Ghannadi-Maragheh
Radiopharmaceutical Research and Development Lab (RRDL), Nuclear Science and Technology Research Institute (NSTRI), Tehran, Iran, Postal code: 14155-1339
166Ho-radiophamaceutical have been developed and used in the therapy of various diseases and malignancies. Holmium-166 microspheres are widely used for the treatment of liver malignancies [1]. Monoclonal antibodies have been radiolabeled using Ho-166 and used in the radioimmunotherapy (RIT) [2]. In this work, anti-CD20 was successively labeled with beta-particle emitting radionuclide, Ho-166, for ultimate radioimmunotherapy applications. Ho-166 chloride was obtained by thermal neutron flux (1 x 1013 n.cm"2.s~') of natural Ho2(NC>3)3 sample, dissolved in acidic media. 166Ho-holmium chloride (185 MBq) was added to the conjugated antibody after ccDTPA residulation at room temperature. Radiochemical purity of 95% (ITLC) and 98% (HPLC) were obtained for final radioimmunoconjugate (Specific activity = 3-3.5 GBq/mg). The final isotonic 166Ho-rituximab complex was checked by gel electrophoresis for protein integrity retention. Biodistribution studies of Ho-166 chloride and radioimmunoconjugate were performed in wild-type rats to determine the biodistribution. The accumulation of the radiolabeled antibody in lungs, liver and spleen demonstrates a similar pattern to the other radiolabeled anti-CD20 immunoconjugates.
1. Zielhuis SW, Nijsen JF, de Roos R (2006) Production of GMP-grade radioactive holmium loaded poly(L-lactic acid) microspheres for clinical application. Int. J. Pharm. 311:69. 2. Dadachova E, Mirzadeh S, Smith SV (1997) radiolabeling antibodies with Ho-166. Appl. Radiat. Isotopes 48:477.
147 NUTECH-2011
PRODUCTION OF FISSION PRODUCT Mo-99 USING HIGH ENRICHED URANIUM PLATES IN RESEARCH REACTOR MARIA THERMAL-HYDRAULIC AND NEUTRONIC ANALYSIS AND TECHNOLOGY
J. Jaroszewicz, Z. Marcinkowska, W. Mieleszczenko, K.Pytel
Institute of Atomie Energy POLATOM, 05-400 Otwock-Świerk, Poland
The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc-99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA Research Reactor. Neutronic calculations and thermal and flow analyses were performed to estimate the fission products activity and heat removal capability for uranium plates which are irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining Mo-99 is based on irradiation of uranium plates in standard reactor fuel channel. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give on- line information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighboring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminum mock- up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed. The activities including technical assumption were focused on performing calculation for modeling of the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminum cladded plate with dimensions of 40x230x1,45 containing 4,7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation.
148 Radiopharmaceuticals and Radioisotope Production
PRODUCTION OF IODINE-125 IN NUCLEAR REACTORS: ADVANTAGES AND DISADVANTAGES OF PRODUCTION IN BATCH OR CONTINUOUS PRODUCTION IN CRYOGENIC SYSTEM
Carlos A. Zeituni1'2, Maria Elisa C. M. Rostelato1, Kwang Jae-Son3, Jun S. Lee3, Osvaldo L. Costa1, João A. Moura1, Anselmo Feher1, Eduardo S. Moura1, Carla D. Souza1, Fabio R. Mattos1, Fernando S. Peleias Jr.1, Dib Karam Jr.4
1 Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN - SP Avenida Professor Lmeu Prestes, 2242 ZIP 05508-000 São Paulo, SP - Brazil. 2 Instituto Presbiteriano Mackenzie, Rua da Consolação, 930 ZIP 01302-907 São Paulo, SP - Brazil 3 Korea Atomic Energy Research institute - KAERI1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353, Korea 4 Escola de Artes, Ciencias e Humanidades - Universidade de São Paulo Avenida Arlindo Bettio, 1000 ZIP 03828-000 São Paulo, SP - Brazil.
Cancer is one of the worst illnesses in the world and one of the major causes of death in Brazil [1,2]. For this reason, the Nuclear Energy National Commission (CNEN) started a project to produce some medical radioisotopes to treat cancer. One of the main products is the iodine-125 seeds [3]. This iodine seed can be used to treat several kinds of cancer: prostate, lung, eye, brain. As Brazil will construct a new reactor to produce radioisotopes, it is necessary define how the iodine-125 production will carry out [4,5]. The main reaction of this production is the irradiation of the enriched xenon-124 in gaseous form. Xe-124 changed to Iodine-125 by neutron capture following in two decays: Xe-124 (n, y) —• Xe-125m (57s) —• I- 125 or Xe-124 (n, y) —• Xe-125 (19.9 h) —• 1-125. However the production in reactors is the most common technique used, there is one disadvantage to use it: the production of iodine- 126 after several hours of irradiation. Iodine-126 has a half life of 13.1 days and it has some usefulness emitters for medical uses. Iodine-126 is considered a contamination [6]. For all these reasons, the IPEN/CNEN-SP research group decided for two techniques of production: in batch or continuous system. The production in batch consists in a sealed capsule that is placed in the reactor core for around 64 hours. In this type of production, some iodine-126 is produced and a certain quantity of Xe-124 is not activated. Normally, it needs to wait around 5 to 7 iodine-126 half-lives to guarantee the decrease of the activity of the contamination. This time will make Iodine-125 with only 50% till 34% of the initial production. The second technique is the continuous production using a cryogenic system. This technique consists in two capsules: one inside the reactor core and the second one out of the neutron flux. These two capsules will be linked with two cryogenic pumps to guarantee that all iodine-125 produced in the core will be take off the reactor core. The great disadvantage of this technique is the using of two positions in the core of the reactor. Brazil will have only one radioisotope reactor producing. And like there is a huge quantity of materials to be produced, it is not a guarantee the position in the reactor for this production. Besides of that the seeds production in Brazil is only 3000 per month, which demands around 3.5 Ci per month. The batch production produces a low quantity per reactor cycle of iodine-125, but this low quantity can be more than that [2,3].
1. MINISTÉRIO DA SA ÚDE. INSTITUTO NACIONAL DE CANCER Incidência de Cancer no Brasil. Available at:
149 NUTECH-2011
2. Rostelato MECM, Rela PR, Zeituni CA, Feher A, Manzoli JE, Moura JA, Moura ES, Silva CPG (2008) Development and production of radioactive sources used for cancer treatment in Brazil. Nukleonika 2008; 53(Supplement 2): 99-S103 3. Moura Ja, Moura Es, Sprenger Fe, Nagatomi Hr, Zeituni Ca, Feher A, Manzoli Je, Souza Cd, Rostelato Mecm (2010) Tubing decontamination during the leak test of iodine- 125 seeds. Nukleonika 2010; 55(3): 409-413. 4. Cieszykowska, I.; Piasecki, A.; Mielcarski, M. (2005) An approach to the preparation of iodine-125 seed-type source. Nukleonika 50; 1:17-22 5. Mathew C, Majali MA, Balakrishnan SA (2002) A novel approach for the adsorption of iodine-125 on silver wire as matrix for brachytherapy source for the treatment of eye and prostate cancer. ApplRadiat Isot 57: 359-367 6. Han HS, Park Ui-J, Son KJ, Lee JS, Hong SB, Ham SS (2007) Development of production technology of 125I seed for brachytherapy. J Label Compd Radiopharm 2007; 50: 321-322
150 Radiopharmaceuticals and Radioisotope Production
PRODUCTION, QUALITY CONTROL AND BIOLOGICAL EVALUATION OF 153Sm-TTHMP AND 153Sm-PDTMP AS A POSSIBLE BONE PALLIATION AGENT
Zohreh Naseri1, Amir Reza Jalilian2*, Ali Nemati Kharat1, Ali Bahrami-Samani2, Mohammad Ghannadi-Maragheh2
1 School of chemistry, University college of science, University of Tehran, Tehran, Iran. Radiopharmaceutical Research and Development Lab (RRDL), Nuclear Science and Technology Research Institute (NSTRI), Tehran, Iran, Postal Code: 14155-1339
Radionuclides have been used to treat skeletal diseases for decades. At present an array of radionuclides have been proposed for treating bone pain due to cancer. The most promising among these radionuclides is 153Sm. This radionuclide has favourable radition characteristics ti/2= 1.93d, ßmax= 0.81 Mev(20%), 0.71 Mev (49%), 0.64 Mev (30%) and y= 103 Kev (30%) emission which is suitable for imaging purposes during therapy[1]. Various bone palliative therapeutic agents have been developed for bone metastasis such as 153Sm-EDTMP, widely used in the world. (EDTMP) has been proved to be a good therapeutic agent for the treatment of pain due to skeletal metastases and is now commercialized. In this study, production, quality control and biodistribution studies of a newly developed therapeutic compound have been presented followed by imaging studies in wild-type rodents. 153Sm-TTHMP and 153Sm-PDTMP was prepared starting 153Sm-SmCl3, prepared by neutron activation of an enriched 152Sm sample (purity >98%), and in-house synthesized TTHMP and PDTMP (amine: 1,2-diaminopropane) in lh at 25°C followed by stability tests, partition coefficient determination and biodistribution studies of in wild-type rodents using scarification. The radiolabeled Sm complex was prepared in high radiochemical purity (>99%, ITLC) and specific activity of 278 GBq/mmol and demonstrated significant stability at 4, 25 and 37°C (in presence of human serum). Initial biodistribution data showed significant bone accumulation of the tracer in 48h. The produced 153Sm-TTHMP and 153Sm-PDTMP properties suggest an efficiently new bone palliative therapeutic agent in the country in order to overcome the metastatic bone pains. Keywords: Sm-153,TTHMP, PDTMP, Therapy, Biodistribution, Radiopharmaceutical.
1. Ouadi A, Loussouarn A, Morandeau L, Remaud P, Faivre-Chauvet A, Webb J, Gestin J-F (2004) Influence of trans-l,2-diaminocyclohexane structure and mixed carboxylic/phosphonic group combinations on samarium-153 chelation capacity and stability. Eur. J. Med. Chem. 39:467-472.
151 NUTECH-2011
THE INFLUENCE OF PARAMETERS OF TARGET MATERIAL ACTIVATION IN A NUCLEAR REACTOR ON EFFICENCY AND QUALITY OF LUTETIUM Lu-177 PRODUCTION
Z. Tymiński, E. Kołakowska, M. Konior, D. Pawlak, A. Patocka
IEA OR POLATOM, Otwock-Świerk, 05-400 Poland
Radioactive contaminations of the artificially produced radioisotopes are typically result of material irradiation in atomic reactor. When the material is produced for medical use, contaminations have to be kept below defined level. The presented results show the impact of the activation parameters of atomic reactor on final product quality. Production efficiency of lutetium 177Lu and its radionuclide purity as a function of time and position in relation to the reactor core during irradiation was examined (Fig. 1). The activity of 177Lu and other radionuclides was measured using a high-resolution gamma ray spectrometer. The developed procedures are used for routine quality control of 177Lu produced for synthesis of radiopharmaceuticals [1]. In additional analysis the standardized activation rate ratio WIRR of 176 LuCl3 target material was applied, which allow the checking of stability of the exposure process. The resulting work constitute the basis for analysis and optimization of the activation process and provide the quality control procedure of the products which is carried out immediately after the activation in the reactor.
10,0E+0 i 0,014
7,5E+0 0,011 d Ï Ś 5,0E+0 fc 0,008 z 0,005 rrr 2,5E+0
0,002 000,0E+0
Production Date Lu-177m [%] Figure. 1. The impurity of 177mLu measured in 177Lu : changes of content in a period of 7 months (left) and the results of measured samples presented in histogram (right).
1. Pawlak D, Parus JL, Sasinowska I, Mikołajczak R (2004) Determination of elemental and radionuclidic impurities in lutetium-177 used for labeling of radiopharmaceuticals. Journal of Radioanalytical and Nuclear Chemistry 261;2:469-472
152 NUTECH-2011
APPLICATION OF NUCLEAR TECHNIQUES FOR MATERIALS SURFACE CHARACTERISATION: OWN INVESTIGATIONS EXAMPLES
B. Sartowska1, J. Piekoszewski1'2, L. Waliś1, W. Starosta1, M. Barlak2 y 3 R Ratajczak , M. Kopcewicz
[email protected] 1 Institute of Nuclear Chemistry and Technology, Warsaw, Poland 2 The Andrzej Soltan Institute for Nuclear Studies, Świerk/Otwock, Poland 3 Institute of Electronic Materials Technology, Warsaw, Poland Knowledge about materials properties has fundamental importance in respect of: (i) material potential application, (ii) possibility of modification - it means improvement - of existing materials, (iii) possibility of production process changes. Different methods and techniques for material characterization are often used as a standard procedure for materials properties determination. Nuclear techniques provide new and more detailed information about investigated materials. Investigations of steels near surface layer modification are carried out in Institute of Nuclear Chemistry and Technology, Warsaw and The Andrzej Soltan Institute for Nuclear Studies, Świerk/Otwock. The main goal of experiments is to improve surface properties including wear, corrosion and high temperature oxidation resistance. Modification processes were carried out using high intensity pulsed plasma beams - HIPPB (106-108 W/cm2) generated in a Rod Plasma Injector (RPI). In most solids such treatment leads to a fast transient melting of the surface layer of the substrate followed by rapid crystallization. Heating and cooling processes were of non-equilibrium type. Two modes of RPI operation are possible: (i) Pulse Implantation Doping - PID when plasma contains practically exclusively ions of the working gas and (ii) Deposition by Pulsed Erosion - DPE when the beam contains also ions/atoms eroded from ends of the electrodes. Near surface layer of the thickness in pm range was melted and simultaneously doped with reactive elements such as nitrogen, rare earth elements (REE). Initial and modified materials were characterised using different investigation methods including nuclear techniques. Examples of nuclear techniques as Nuclear Reaction Analysis (NRA), Rutherford Backcscattered Spectroscopy (RBS) and Mössbauer spectroscopy are described in the paper. Results of modified surface characterization using standard and more informative nuclear methods are presented in the paper.
154 CharacterizaƟon of Materials
APPLICATION OF XRF AND GC-SYSTEM FOR SOURCE IDENTIFICATION OF ARCHEOLOGICAL SAMPLES FOUNDED IN MENTASHTEPE OF AZERBAIJAN D. Abbasova [email protected]
Institute of Radiation Problems of AzNAS, Azerbaijan and use oils and oil products for the different purposes. Bitumen properties were well known from an as preventive materials for the wood balks and also for the art. This was also approved by the samples founded on the Azerbaijan territory in summer 2009. During the excavation provided by Dr. F. Guliyev in Menteshtepe (Azerbaijan NAS) and Bertille Lyonet (France) of Tovuz region in Azerbaijan were founded ceramics with the bitumen inside. The samples were analyzed by radiocarbon dating method on AMS in CNRS and dated by 3500BC. The most interesting fact attracted us was that there are no any oilfields situated there or ever has been. That is why the analysis of bitumen was very important and could help to find out the origin of this bitumen. By biomarkers analysis can be found place where this ceramics was taken and in conclusion will be helpful to suppose the possible historical movement of settlement. For identification of source of bitumen origin has been used chromatographic methods. Sample was analyzed for the Total Petroleum Hydrocarbon (TPH) including Pristan and Phythan and metal contents by newest equipments GC and XRF (USA). The ratio of Pr/Ph was less than 1 (is 0.04), where phytane in domination, which indicates that our sample belong to Saltwater facies and hyper-saline sedimentary rocks. Low values of the ratio Pr/Ph and Pr/n-C17 is typical for offshore deep-water facies, as well as evidence of the original sapropel organic matter. Increased phytane is an indicator of reducing environment at an early stage of diagenesis. Was also identified that bitumen samples belong to Paleocene period. The metal Al >Zn>Cu>Ni>As>Pb>Sn correlation between bitumen from different oilfields shows that sample is belong to Bandovan oilfields bitumens of Azerbijan. Conclusion: Organic analysis help to identify the organic matter origin by using the biomarkers and metal contents shows the correlation with bitumen with such type of characteristic metal content. Was founded that sample was belong to Bandovan oilfields of Azerbaijan.
155
NUTECH-2011
DETERMINATION OF URANIUM (VI) AND THORIUM (IV) IN TECHNOLOGICAL PHOSPHORIC ACID SOLUTIONS
Z. Samczyński
Institute of Nuclear Chemistry and Technology, ul. Dorodna 16, 03-195 Warszawa, POLAND
A method for the determination of uranium (VI) and thorium (IV) by spectrophotometry in technological phosphoric acid solutions has been elaborated. It is based on the complex forming reaction between the chromogenic reagent Arsenazo III and U (VI) as well as Th (IV) [1]. Their complexes show the maximum absorbance in the visible region at 655 nm. Phosphate ions, however, strongly interfere with the reaction. For this reason, direct determination of U and Th in phosphoric acid medium is not possible and thus their isolation is necessary. Ion exchange chromatography was applied for this purpose, employing amphoteric ion exchange resin Retardion 11A8. It contains strongly basic ammonium groups and an equivalent amount of weakly acidic carboxylate groups. Depending on the composition of the external solution, the anion exchange (ammonium) groups can retain anions as well as the cation exchange (carboxylic) groups can take up cations. Uranium (VI) and thorium (IV) in phosphoric acid solutions exist in the form of anionic phosphate complexes, which can be sorbed by the ammonium groups of Retardion 11A8. The resin acts as an anion exchanger in these conditions. The highest affinity of U and Th towards the resin is observed in diluted solutions of H3PO4. With the increasing of the concentration of the acid, their sorption on Retardion 11A8 rapidly diminishes. Phosphoric acid solution of the concentration approx. 0.5 mol L-1 has been established as optimum for the isolation and preconcentration of U (VI) and Th (IV). The next stage of the ion exchange procedure is the separation of thorium from uranium, performing elution with 0.1 mol L-1 solution of sodium salt of nitrilotriacetic acid of pH approx. 6.80. Uranium stays firmly fixed on the column during this process, while thorium is washed out. In nearly neutral solutions (pH 6-8), carboxylic groups of Retardion 11A8 are capable of retaining cations and the resin can act as a cation exchanger as well. As can be inferred, in this medium, U and Th are bound on the stationary phase by carboxylic groups in the form of positively charged species. Their different stability underlies the separation of the analytes. Finally, U (VI) is stripped from the column by means of 0.5 mol L-1 HCl solution.
Accuracy of the devised method was examined by adding known amounts of uranium and thorium into pure (of analytical grade) solution of phosphoric acid. The recovery amounted to 100% ± 3%. The content of U and Th was analyzed in real samples, taken two different batches of technological phosphoric acid, produced by processing separately phosphate ores imported from Morocco and Tunisia. The determined content of uranium in H3PO4 amounted to 210 and 125 [j,g mL"1, respectively. In the both cases, thorium was not found.
Acknowledgement: This work was supported by project POIG.Ol.01.02-14-094/09, "Analysis of the possibility of uranium supply from domestic resources " 1. Marczenko Z., Spectrophotometric Determination of Elements, Ellis Horwood, Chichester, 1986.
156 Characterization of Materials
DETERMINING THE CONTENT OF 10B IN BORIC ACID BY MEANS OF THE THERMAL NEUTRON ABSORPTION TECHNIQUE
A. Bolewski Jr, M. Ciechanowski, A. Kreft
AGH University o Science and Technology, Faculty of Physics and Applied Computer Science, al. Mickiewicza 30, 30-059 Krakow, Poland
Boron and its compounds are extensively used in nuclear industry as strong thermal neutron absorbers. In particular, boric acid (H3BO3) is added to the primary circuit coolant of pressurized water reactors (PWR) to control the chain reaction. Owing to the application of the neutron-absorbing coolant, the irregularities in the power density distribution and fuel consumption within the reactor core can be avoided, with the result that performance of nuclear power plant can be improved. Natural boron is a mixture of 10B and 11B isotopes with abundances of about 20 % and 80 %, respectively. Thermal neutron absorption cross sections of 10B and 11B are 3839 b and 0.0055 b, respectively. Therefore, the actual factor affecting the reactor performance is the concentration of 10B isotope in the coolant. A continuous monitoring of that concentration is of great importance. Uncontrolled increase or reduction of 10B content would lead to an operational upset. It is known that the isotopic composition of natural boron varies significantly depending on the origin of the raw material. Furthermore, 10B added to the reactor coolant is steadily burnt out during the reactor operation. For those reasons chemical methods are not adequate for determining the content of 10B of in the reactor coolant. This resulted in motivation for the development of measurement systems sensitive specifically to 10B isotope. Techniques based on the thermal neutron absorption appeared to be especially suitable for this purpose. Different versions of such devices designed for continuous on-line or off-line monitoring the concentration of 10B in the reactor coolant have been developed and perfected since the beginning of the PWR technology. All instruments of this type require calibration. For the sake of metrological traceability of the measurement system, any calibration procedure should finally refer to a certified boric acid isotopic reference material. However, because of high cost, the isotopic reference materials are seldom, if ever, used directly for preparation of calibration solutions. Usually some secondary standard of boric acid is used for this purpose. The aim of this work was to improve with the aid of MCNP modelling the thermal neutron absorption technique developed earlier and to apply it to determining the content of 10B in boric acid. A good deal of attention was given to optimizing the measurement set up and procedure in order to reduce an uncertainty of assays. With the use of 1370-ml sample, 252Cf neutron source emitting about 2 x 107 neutrons/s and assuming 60-minute total counting time, the relative standard deviation of 0.12 % can be attained for determining the mass fraction of 10B in water solution of boric acid in the range 160-750 ppm. Through the measurement of the mass fraction of 10B in an intentionally prepared water 10 solution of boric acid (C) the mass fraction of B in this boric acid (CBA) can be easily determined. Namely, CBA= C/CA, where CA is the mass fraction of boric acid in its water solution. If boric acid is of stoichiometric purity, the isotopic composition of boron can be derived from CBA . Owing to good precision and reliability the presented technique is suitable for preparing secondary standards for 10B assays and could be helpful in nuclear plants with pressurized water reactors.
157 NUTECH-2011
EFFECTIVENESS AND LIMITATIONS OF QUANTITATIVE NEUTRON IMAGING CORRECTIONS FOR ACCURATE CHARACTERISATION OF POROUS MEDIA
M. J. Radebe1, F. C. de Beer1, R. B. Nshimirimana1, J. J. Milczarek2,1. M. Fijał-Kirejczyk2, J. Żołądek-Nowak2, G. Nothnagel1
Radiation Science, Necsa, Church street west ext., Pelindaba, Pretoria, 0001, South Africa institute of Atomic Energy POL ATOM, 05-400 Otwock-Świerk, Poland
Characterization of the properties of porous media and/or its validation is very important in nuclear liabilities management in the context of containment barriers and encapsulation of nuclear waste in media such as concrete or related materials. Digital neutron radiographs contain inherent density information relating to the radiation attenuation by the object. It therefore offers the possibility of quantitative investigation of the amount and distribution of a particular embedded material and particular physical properties related to it as a function of position in the object through tomography application. Accurate and precise quantitative measurements are only possible when appropriate corrections for beam energy spectrum, neutron attenuation and neutronscattering are made. The latter includes scattering form the test object as well as from the surrounding environment. In addition the detector response properties need to be properly accounted for. Efforts have been made towards establishing and implementation of these corrections by a number of researchers [1,2,3]. However, the quality of the scatter point-spread-functions (PSF) used in the correction of material scatter, as well as alternatives such as COG code which has been established for neutron imaging of complex geometries of material, have not been examined nor explored[4]. This research seeks to evaluate and to improve on these quantitative correction measures when applied in e.g. water calibration samples, amount of water generated in fuel cells and in porous media with free water in connected pores. The experiments were carried out at the SAFARI-1- and MARIA nuclear research reactors situated in South Africa and Poland respectively.
1. Kardjilov N, De Beer FC, Hassanein R, Lehmann E, Vontobel P. (2005) Scattering corrections in neutron radiography using point scattered functions. Nucl. Instr. Meth. Phys. Res. A 542. pp. 336-34J. 2. Hassanein R, De Beer F.C., Kardjilov N., Lehmann E. (2006) Scattering correction algorithm for neutron radiography and tomography tested at facilities with different beam characteristics. Physica. B 385-386. pp. 1194 1196. 3. Hassanein R Lehmann E, Vontobel P (2005) Methods of scattering corrections for quantitative neutron radiography. Nucl. Instr. Meth. Phys. Res. A 542. pp. 353-360. 4. Wilcox T, Lent E (2002) COG User 's Manual. Fifth Edition.
158 Characterization of Materials
INSTRUMENTAL NEUTRON ACTIVATION ANALYSIS (INAA) FOR STEEL ANALYSIS AND CERTIFICATION.
E. Chajduk, B. Danko, H. Polkowska-Motrenko
Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland
Addition of trace elements (intended or as impurities) in steel or alloys can completely change the properties of materials. Thus, there is necessity to have reliable analytical techniques to determine major and trace components in this type of material. Also, to solving problems with quality control/ quality assurance (QC/QA) in this subject, appropriate certified reference materials should be available in the wide range. Several methods have been proposed for quantitative analysis of steel samples: XRF, ICP-OES, ICP-MS, AAS and NAA. INAA is one of the preferred methods because it provides information on a large number of elements simultaneously. Another important feature of INAA is that it is free from the problems connected with digestion of difficult matrices and potential losses of elements and/or contamination of a sample from reagents. Moreover, by changing irradiation and cooling parameters as well as measurement time, the effect of the interfering radionuclides on the results of analysis of particular elements can be decreased. Because of these advantages, INAA is used as microanalytical method in metallic material analysis. During recent years it has been proved that instrumental NAA can meet CCQM criteria for a primary method of measurement (PMM). Such methods are applied for certification of reference materials, because they establish metrological treaceability. That is why INAA was used for certification of CRM of steel origin. INAA procedure including irradiation conditions, decay and measurement times, interpretation of the results for the elemental analysis of metallurgical samples has been elaborated and presented. The results of steel analysis were compared with the results obtained by classical analytical techniques. It has been proven that procedure gives accurate results for several elements at different concentration range with very low combined uncertainty. The results obtained when participated in Proficiency Testing/Interlaboratory Comparison (PT/ILC) exercises concerning elemental analysis of steel material by neutron activation analysis are presented.
Acknowledgement: The authors would like to thank Dr Em. Cincu from IFIN-HH, Bucharest -Magurele, Romania for organisation of PTs and providing steel samples.
159 NUTECH-2011
METHOD FOR DETECTION OF HIDDEN EXPLOSIVES AND OTHER ILLICIT MATERIALS WITH USE OF NANOSECOND NEUTRON PULSES FROM PLASMA FOCUS DEVICE - MONTE CARLO MODELLING OF THE SYSTEM
U. Wiącek1, R. Miklaszewski2, V.A. Gribkov3, B. Gabańska1
lrThe Henryk Niewodniczański Institute of Nuclear Physics Polish Academy of Sciences, Kraków, Poland 2Institute of Plasma Physics and Laser Microfusion, Warsaw, Poland 3Institute of Theoretical and Experimental Physics, Moscow, Russia
Monte Carlo simulations of fast neutron scattering for a development of a single-pulse Nanosecond Impulse Neutron Investigation System (NINIS) are presented. The NINIS is a new method dedicated for detection of hidden illicit objects (e.g. explosives, fissile materials, etc.). After irradiation of an interrogated object with a pulse of fast neutrons, it is possible to determine the elemental content of unknown samples from information existing in a field of scattered neutrons. The method needs very short and intense neutron pulses having duration of the order of a few nanoseconds. Such conditions can be fulfilled by a dense plasma focus device. A very short duration (~10 ns) of the neutron pulse, mono-chromaticity and a high intensity (up to 109 of 2.45 MeV neutrons per shot from the D-D reaction or up to 1011 of 14 MeV neutrons from the D-T reaction) allow to use the time-of-flight method with bases of about few meters to distinguish signals from neutrons scattered by different elements. Results of the Monte Carlo modelling of the scattered neutron field from several compounds (explosives and everyday use materials) are presented. The MCNP5 code was used to get info on the angular and energy distributions of the scattered neutrons. A plasma focus source of the D-D neutrons was assumed for irradiation of investigated objects. Neutron collisions and transport in the objects were watched and registered in time and energy up to the moment of detection in an assigned position. The simulations were performed for various variants, like: various materials (illegal and everyday use materials), various distances from the scattering sample to the detector, various angles of detection. The final MCNP input allows modelling performance of the whole detecting system while inspecting an exemplary air luggage. The obtained results show influence of presence of various materials in the luggage on registered signals and demonstrate a principal capability of the method to identify an elemental content of the inspected objects.
160 Characterization of Materials
QUANTITATIVE CHARACTERIZATION OF STRATIFIED MATERIALS BY CONFOCAL 3D MICRO-BEAM X-RAY FLUORESCENCE SPECTROSCOPY
M. Czyzycki, D. Wegrzynek, P. Wrobel, M. Lankosz
AGH University of Science and Technology, Al. Mickiewicza 30, PL-30059 Cracow, Poland
Components based on multi-layered systems play currently a meaningful role not only in surface and material technology, but in almost all branches of modern industry. Materials with stratified structure are very attractive for practical applications in electronics (dielectric coatings for superconductor devices) as well as energy production (solar energy cells) and energy storage (rechargeable lithium batteries). Some accessories of optics (optical fibre) consist of multi-layered elements. Stratified materials are generally applied as environmental solutions (functional membranes for gas separation, gas chromatography, air and water purification, nuclear waste treatment) and in petroleum industry (catalytic cracking of fuel). Multi-layered materials are commonly utilized in automotive industry (lacquer systems) and in building industry (ceramics). Stratified materials are often met as components in biomedical applications (biosensors, biocatalysts and biopharmaceuticals). Hence, elemental analysis of these materials is of great interest. Examination of chemical composition of individual layers and determination of their thickness helps to get information on their properties and function. A confocal 3D micro X- ray fluorescence (3D pXRF) spectroscopy is an analytical method giving the possibility to investigate 3D distribution of chemical elements within a sample with spatial resolution in micrometer regime in non-destructive way. This paper demonstrates the practical application of the Monte Carlo (MC) simulation of the 3D ¡iXRF experiment performed on multi-layered materials. The stratified materials with thickness up to 500 |im based on low-Z polymer matrix doped with small quantities of oxides were examined with synchrotron radiation as test samples. Elemental concentrations and thickness determined with the MC simulation were compared against predictions of a fundamental parameter model proposed to 3D pXRF of multi-layered samples.
161 NUTECH-2011
THERMAL NEUTRON RADIOGRAPHY STUDIES OF DRYING OF RECTANGULAR BLOCKS OF WET MORTAR
I. M. Fijal-Kirejczyk1, J. J. Milczarek1, F. C. de Beer2, M. J. Radebe2, J. Żołądek-Nowak1
institute of Atomic Energy POLATOM, Świerk, 05-400 Otwock, Poland, 2 Radiation Science, Necsa, Church street west ext., Pelindaba, Pretoria, 0001, South Africa
The drying process of capillary-porous materials is of great commercial as well as basic phenomena interest. The sensitivity of neutron imaging to presence of water is a main merit of the neutron radiography in studies of all phases of drying. Our previous studies dealt with drying of cylinders made of various porous materials like quartz sand and kaolin clay [1, 2]. In this work the drying of rectangular block made of mortar and saturated with water was studied. The experiment was carried out at the neutron radiography facility NGRS at the MARIA neutron research reactor. The sample was dried in the drying tunnel with 65°C hot air stream. The sequences of thermal neutron radiographs were collected simultaneously with the sample temperature and mass measurements. The constant rate period (CRP) of drying, the falling rate period (FRP), as well as receding front phase was identified on the basis of analysis of recorded neutron radiographs. The sample was a rectangular block of 52 mm x 48 mm x 14 mm made of aged mortar which was saturated with water by immersion for 48 h. The same sample was investigated first with all its surfaces free and then after painting its four sides with top and bottom surfaces left uncovered with paint. The main differences between results obtained in these two different arrangements consisted in lower drying rate and narrowing of the inner wet region contained inside receding front surface for the painted sample in comparison to unpainted one. The analysis of the thermal neutron radiographs was performed within the previously developed method [2] and revealed the increase of the average brightness of the sample image and a well developed maximum in the plot of the standard brightness deviation vs time. The last feature is in contrast with findings for drying of cylinders where two maxima in standard deviation time dependence were found. We found that the maximum of brightness standard deviation indicates the most developed inner wet region inside the sample. The successful application of quantitative thermal neutron radiography in studies of drying of rectangular samples supports this technique as a versatile tool for identification and characterization of all phases of drying for objects of different shapes.
1. Fijał-Kirejczyk IM, Milczarek .J.J, Banaszak J, Trzciński A, Żołądek J (2009) Dynamic neutron radiography studies of drying of kaolin clay cylinders. Nukleonika 54;2: 123 - 128 2. Fijał-Kirejczyk IM, Milczarek .J.J, Banaszak J, Trzciński A, Żołądek J (2011) Neutron radiography observations of inner wet region in drying of quartz sand cylinder. Nucl. Instrum. Methods. A. http://dx.doi.org/10.1016/j.nima.2011.02.037
162 NUTECH-2011
BIOKINETICS AND RADIATION DOSIMETRY FOR [4-14C]-CHOLESTEROL IN HUMANS
L. A. Marcato, M. M. Hamada and C. H. Mesquita
Energy and Nuclear Research Institute, CNEN/SP, São Paulo, Brazil
Medical and clinical researches utilize radiolabelled cholesterol to obtain information about the physiology of cholesterol and of its several substrates (biliary acids, hormones and vitamins) in the body. The radiotracers constitute a simple and accurate tool for metabolic studies; however, the scientific community has shown certain reservations concerning the use of radioisotopes. Probably, the apprehension is result of the question about the deleterious radiation effects. Although the studies that utilize radioisotopes are approved by strict ethic committees, most of them do not mention the radiometric doses at which the human subjects are exposed during these studies. The International Commission on Radiological Protection (ICRP) provides a generic carbon model (GCM) to calculate the effective dose of compounds labeled with 14C, first described on ICRP publication 30. The effective dose coefficients for most compounds appear to be greatly overestimated by the GCM in comparison with those generated by more realistic models [1]. The GCM cannot be applied to the interpretation of bioassay data with any degree of confidence [1]. The purpose of the present study is to improve the generic biokinetic model [2] for use in the assessment of the internal dose received by human subjects who were administered labelled cholesterol either orally or intravenously. This model was used with the ANACOMP software to estimate the radiometric doses with the MIRD techniques. To validate the model, the simulated profile curves were compared with the profile curves described on the literature (Kruskal-Wallis test, P=0.4232). The model reproduced the intestinal absorption of cholesterol and the excretion of cholesterol in feces and urine. The estimated effective dose coefficient calculated for the reference man described on ICRP publication 23 was 1.35x10-11 SvBq-1. The organs that received the highest equivalent dose were the lower large intestine (1.03x10-10 GyBq-1), upper large intestine (3.74x10-11 GyBq-1) and small intestine (1.58x10-11GyBq-1). The effective dose coefficient calculated by the proposed dosimetric model was approximately forty-three times lower than that which is calculated by the ICRP generic model (5.8x10-10 SvBq-1) for ingested 14C that assumes complete absorption to blood.
1. Taylor DM (2004) Bioknetic models for the behavior of carbon-14 from labelled compounds in the human body: Can a single generic model be justified? Radiation Protection Dosimetry 108;3:187-202 2. Manger RP (2011) A generic biokinetic model for cabor-14. Radiation Protection Dosimetry 143;1:42-51
164 Biomedical Studies
ELEMENTAL QUANTIFICATION OF THIN TISSUE SAMPLES IN SR-XRF TECHNIQUE USING EXTERNAL STANDARDS
M. Szczerbowska-Boruchowska
AGH University of Science and Technology, Faculty of Physics and Applied Computer Science, Krakow, Poland
Synchrotron radiation based x-ray fluorescence analysis (SR-XRF) offers a nondestructive qualitative and quantitative analysis of trace elements. High photon flux of the incident beam and high spatial resolution allow for the mapping of minor and trace element distributions in biological samples at the cellular scale [1,2]. Since the elemental study is carried out for the intact parts of the tissue or cells the external standard method remains the only way of quantitative evaluation. The main problem that occurs in this field is selection of appropriate standard samples. A typical thickness of the specimens studied using SR-XRF technique is about 1-^2 |iin therefore thin film approximation that neglects absorption effects of the exciting and the detected radiation is valid. The standard samples of high elemental homogeneity are crucial taking into account the beam size applied in the analysis. Some thin standard reference materials are commercially available (e.g. NIST SRM 1832, NIST SRM 1833, or MICROMATTERTM XRF calibration standards from). However they allow mainly for the determination of masses per unit area of elements. The standard reference materials of biological origin for which the certified values of elemental concentrations are provided typically occur in lyophilized powder form. Therefore they are not appropriate for microprobe analyses. In the presented study to determine elemental mass fractions in biological specimens the home made standard samples were applied. For this purpose a mixture of nitrates of metals such as Cl, Ti, K, Fe, Sc, Zn, Mn, Rb, Cu, Y, Se, Sr in Tissue Freezing Medium (Jung Leica) were prepared. Two kinds of standard samples of different elemental composition were made. The specimen was frozen at -20 °C and cut into sections with a thickness of 20 pm using a cryomicrotome. The sections were placed on ultralene foil stretched on Plexiglas discs and dried. The SR-XRF measurements were carried out at the 7T-WLS/1 (mySpot) beamline of the Electron Storage Ring BESSY II (Berlin, Germany). The primary photon energy was set to 17 keV. Capillary optics was used to focus the X-ray beam on the sample surface to spot sizes of about 12 pm in diameter. The homogeneity of the standard samples was verified in two slices of each specimen. 49 different points of each sample were probed. The results show that the relative standard deviations calculated for fluorescence intensities of measured elements ranging from 0.5% to 5.2%. High homogeneity and accuracy of the thin standard samples was achieved.
1. Ortega R, Bohic S, Tucoulou R, Somogyi A, Devès G (2004) Micro-chemical element imaging of yeast and human cells using synchrotron X-ray microprobe with Kirkpatrick- Baez optics. Anal. Chem. 76:309-314 2. Twining BS, Baines SB, Fisher NS, Maser J, Vogt S, Jacobsen C, Tovar-Sanchez A, Sañudo-Wilhelmy SA (2003) Quantifying trace elements in individual aquatic protist cells with a synchrotron X-ray fluorescence microprobe. Anal. Chem. 75:3806-3816.
165 NUTECH-2011
EXOGENOUS OR ONTOGENETIC FACTORS INFLUENCE ON CELLULAR RADIOSENSITIVITY AND CANCER INCIDENCE IN THE POPULATION MONITORED WITH BIOMARKERS APPLIED FOR THE RETROSPECTIVE BIOLOGICAL DOSIMETRY
A. Cebulska-Wasilewska
Institute of Nuclear Physics PAN, Department of Radiation and Environmental Biology, Radzikowskiego 152, Krakow, Poland
The aim of our human monitoring studies performed on subjects from various cities and countries was to investigate if emergency or occupational exposures to genotoxic agents can cause a detectable health risk. Exposures to ionizing radiation, pesticides, mercury ions, benzene related compounds have significantly elevated levels of cytogenetic damage. The 25 years follow up studies revealed significantly increased risk of cancer in the group of subjects characterized by the highest levels of the detected chromosomal damage. Results of DNA repair competence assay, with a use of challenging dose of X-rays and the detection of induced DNA damage by the SCGE assay, have correlated to levels of induced chromosome damage. Results from studies on the influence of occupational exposure to PAHs on the repair of DNA damage induced by radiation, have shown a strong variability between donors and significant decrease of the DNA repair efficiency in exposed subjects, that was strongly differentiated between groups stratified first according to various genotypes for genes, encoding enzymes involved in the process of bio-transformation (CYP1A1(Ile/Val), GSTM1, NAT2) or DNA repair (EPHX4 or XRCC1)1 and then to levels of exposures. Results of our studies have also shown the higher levels of chromosome aberrations frequencies and associated significant reduction of cellular repair efficacy, that were observed in a various groups of cancer patients when compared to healthy subjects. Presented results point towards the DNA repair competence biomarker as well as ontogenetic or exogenous factors which via alteration of the DNA repair processes can rise levels of chromosome aberrations and result in increased health risk, that knowledge might be key factor in the ranking list for stratification of the population at risk of mass casualty.
Acknowledgments: Research was partially supported by grant: "Studies on individual susceptibility to radiation- induced damages, DNA repair and instability of human chromosomes in lymphocytes of patients diagnosed or treated with radioiodine" (0296/B/P01/2008/3 5) and "Estimation of individual radiosensitivity of patients with prostate cancer and benign prostatic hyperplasia with application of FISH technique" (2520/B/P01/2010/39).
1. Cebulska-Wasilewska A, Binkova B, Sram RJ, Kalina I, Popov T, Farmer PB, (2007) Repair competence assay in studies of the influence of environmental exposure to c-PAHs on individual susceptibility to induction of DNA damage. Mutat. Res./Fundamental and Molecular Mechanisms of Mutagenesis 620, 1-2, 1, 155-164.
166 Biomedical Studies
INVESTIGATION OF IRON AND ZINC OXIDATION STATE IN DIFFERENT GRADES OF HUMAN BRAIN GLIOMAS USING XAFS SPECTROSCOPY
K. Wolska1, A. Wandzilak1, M. Czyżycki1, P. Wróbel1, M. Szczerbowska-Boruchowska1, D. Adamek2, E. Radwańska3, M. Lankosz1
1Faculty of Physics and Applied Computer Science, AGH University of Science and Technology, Kraków, Poland 2Department of Neuropathology, Chair of Pathomorphology, Faculty of Medicine, Jagiellonian University, Botaniczna 3, 31-503 Krakow, Poland ^Department of Neuropathology, Institute of Neurology, Faculty of Medicine, Jagiellonian University, Botaniczna 3, 31-503 Krakow, Poland
Brain glioma is the most common form of brain cancer. The aim of this study was to investigate iron and zinc oxidation state of different grades of human brain gliomas. Seven different samples of brain tissue were examined using X-ray absorption near-edge structure (XANES) and extended X-ray absorption fine structure (EXAFS) methods. We were investigated one abscess sample and samples of II, III and IV grade gliomas (two of each grade). The position of zinc absorption edge suggested that in all samples zinc exist in bounded form Zn (II). Position of iron absorption edge suggested than gliomas tissue contain mixture of Fe (II) and Fe (III). It was noticed about 1.2-1.5 eV shift between iron absorption edge for II grade gliomas and for III or IV grade gliomas. This difference suggested higher Fe (II) content in III or IV grade gliomas than in II grade gliomas. Iron absorption edge for abscess sample were very similar to absorption edge in Fe2O3 sample, used as a standard sample. Preliminary EXAFS studies showed small differences between spectra for iron absorption edge for samples of different grade gliomas. This could suggest change in distance to neighboring atoms between different samples.
Acknowledgments: This work was supported by the following grants: the Ministry of Science and High Education, Warsaw, Poland (N N518 377537), HASYLAB experimental grants (1-20090047 EC and 1-20100040 EC) and the European Community's Seventh Framework Programme (FP7/2007-2013) under grant agreement n° 226716.
1. Al-Ebraheem A, Goettlicher J, Geraki K, Ralph S, Farquharson MJ (2010) The determination of zinc, copper an iron oxidation state in invasive ductal carcinoma of breast tissue and normal surrounding tissue using XANES. X-RAY Spectrometry 2010, 39, 332-337 2. Szczerbowska-Boruchowska M (2008) X-ray fluorescence spectrometry, an analytical tool in neurochemical research. X-RAY Spectrometry 2008, 37, 21-31 3. Yoshida S, Ide-Ektessabi A, Fujisawa S (2003) Application of Synchrotron Radiation in Neuromicrobiology: Role of Iron in Parkinson 's Disease. Structural Chemistry, Vol. 14, No. 1 4. Huang X (2003) Iron overload and its association with cancer risk in humans: evidence for iron as a cancerogenic metal. Mutation Research 533 (2003) 153-171 5. Ke Y, Qian ZM (2007) Brain iron metabolism: Neurobiology and neurochemistry. Progress in Neurobiology 83 (2007) 149-173
167 NUTECH-2011
MEASURING THE LEVELS OF SOME TRACE ELEMENTS IN THE BLOOD OF PATIENTS SUFFERING FROM MULTIPLE SCLEROSIS USING NEUTRON ACTIVATION ANALYSIS
M.N. Nasrabadi1, D. Forghani
Department of Nuclear Engineering, Faculty of Advanced Sciences & Technologies, University of Isfahan, Isfahan 81746-73441, Iran
Multiple Sclerosis (MS) is a neurological autoimmune disease in which the immune system attacks the central nervous system for unknown reasons and causes several damages to human body by demyelinating the nerve cells. One of the possible causes of this disease is the abnormality of trace elements levels in human body. Neutron activation analysis is one of the most precise methods for determining trace elements in blood. This study attempts to measure the level of four trace elements of Bromine (Br), Ferrum (Fe), Rubidium (Rb), and Zinc (Zn) in the patients' blood samples and compare them with control samples from healthy individuals. According to the obtained results, the differences between the levels of Br, Fe, and Rb in patients' blood samples and control was not significant (P> 0.05). However, the average level of Zn between samples and controls showed a significant difference (P< 0.05). Also it was observed that the lower level of Zn in blood can be a major cause of MS emergence. Furthermore, it was revealed that the risk of MS infection rises as the number of pregnancies increases.
1. Abramovitz M, (2002) Multiple Sclerosis, Lucent Books 2. Ehmann WD, Vance DE, (1991) Radiochemistry and Nuclear Methods of Analysis, Wiley, New York 3. Alfassi ZB, (1994) Chemical Analysis by Nuclear Methods, Wiley, Chichester 4. Schulten HR, Palavinskas R, Kriesten K, (1983) Time-Dependent Excretion of Lithium, Sodium, Potassium, Rubidium, Magnesium and Strontium in the Urine of a Multiple Sclerosis Patient, BiomedMass Spectrom. Mar;10(3):192-196 5. Palm R, Hallmans G, (1982) Zinc and Copper in Multiple Sclerosis, Journal of Neurology, Neurosurgery, and Psychiatry 45; 691-698 6. Visconti A, Cotichini R, Cannoni S, et al., (2005) Concentration of Elements in Serum of Patients Affected by Multiple Sclerosis with First Demyelinating Episode: a Six-Month Longitudinal Follow-Up Study, Ann Ist Super Sanità 41 : 2; 217-222
168 Biomedical Studies
POTENTIALITIES OF 109CD-BASED X-RAY FLUORESCENCE FOR IN VIVO LEAD CONTENT IN BONE
M. L. Carvalho1, Eric Da Silva2,J- P. Marques1, M.T. Lima1, A. Pejovic-Milic2'3, D. R. Chettle2
dentro de Física Atómica e Departamento de Física da Universidade de Lisboa Av. Prof. Gama Pinto 2, 1649- 003, Lisboa, Portugal 2Medical Physics and Applied Radiation Sciences Department, McMaster University, Hamilton, Ontario, L8S 4K1, Canada 3Department of Physics, Ryerson University, Faculty of Engineering, Architecture and Science, 350 Victoria Street, Toronto, Ontario, M5B 2K3, Canada
Lead is a heavy metal with hazardous effects on the human body, even at low concentration levels. Due to its high affinity to calcium binding sites, lead can accumulate especially in bone and nervous system. The in vivo/in situ measurement of bone lead by K XRF has become common in the evaluation of lead metabolism and in its application to epidemiological studies [1]. The XRF system has also been exploited for its portable and non-destructive in situ capabilities for its general application to archeological studies in which bone lead concentrations are of interest [2, 3]. The XRF system is portable in nature and the general procedure and physical methodology being based on the detection of the Pb Ka and Kß lines. The system consists of a HPGe germanium detector equipped with a 109Cd excitation source. Samples are measured in the 180° backscatter geometry relative to the excitation source which allows for Pb signal normalization via the coherent scattering of the 109Cd -ray, allowing for correction of differences in bone size and volume, and in the case of in vivo studies, allows for overlying soft tissue correction. In the current method, this is accomplished by means of calibration using Plaster of Paris standards with known concentrations of lead from 1-200 |ig/g.
1. Hu, H., Milder, F.L. and Berger, D.E., Environmental Health Perspectives, 94, 107-110 (1991). 2. Rebôcho, J., Carvalho, M.L., Marques, A.F., Ferreira, F.R. and Chettle, D.R, Talanta, 70, 957-961 (2006). 3. Keenleyside, A., Song, X., Chettle, D.R. and Webber, C.E. Journal of Archaeological Science, 23, 461-465 (1996).
169 NUTECH-2011
RADIOEMBOLIZATION OF COLORECTAL CANCER HEPATIC METASTASES USING YTTRIUM-90 MICROSPHERES - PRELIMINARY REPORT.
Z. Podgajny 1 , P. Piasecki 2 , N. Szaluś 3, P. Zięcin• • a 2, J. Barzał 4 , K. Brzozowski 2 , G. Kamińsk• i• 1
Departments of Endocrinology and Radioisotope Therapy1, Radiology2, Nuclear Medicine3, Oncology4 Military Institute of Medicine, Warsaw, Poland
Background: intrahepatic arterial yttrium 90(90Y) microspheres have been proposed as a less toxic, less invasive therapeutic option to chemotherapy or transhepatic arterial chemoembolization for patients with surgical unresectable hepatic metastases of colorectal cancer. Aim: The objective of the current study was to determine the safety and efficacy of 90Y microspheres treatment in patients with hepatic metastases of colorectal cancer. Material and Methods: since June 2009 to March 2011 eleven patients with unresectable hepatic colorectal cancer metastases have been treated with radioembolization. 90Y microspheres (SIR-Sphers) were used in dose of : 0.59GBq - 1.93 GBq (mean 1.37GBq). Safety and toxicity were assessed according to National Cancer Institute Common Terminology Criteria version 3 (NCICTC v.3). Metabolic response was assessed by 18F-FDG positron emission tomography and computed tomography (PET-CT). Results: we obtained complete metabolic response (CR) in two cases (18%), partial metabolic response (PR) in six cases (55%), metabolic stabilization disease (SD) in one case (9%) and metabolic progression disease (PD) (18%) in two cases. There were observed treatment- related toxicities: fatigue, nausea and abdominal pain, only grade 1 and 2 according to NCICTC v.3. The median time to hepatic progression of metastases and overall survival will be assessed after completing the study. Conclusions: Y-90 liver radioembolization therapy appears to provide to metabolic response of disease with acceptable toxicity. Key words: radioembolization, SIR-Sphers, colorectal cancer, hepatic metastases
170 Biomedical Studies
SPECT AND PET IMAGING OF YTTRIUM-90
A. Budzyńska1, N. Szaluś1, E. Dziuk1, M. Dziuk1, G. Kamiński1, D. Pawlak2
1Military Institute of Medicine, Warsaw, Poland institute of Atomic Energy POL ATOM Radioisotope Centre, Świerk-Otwock, Poland
Purpose: Yttrium-90 (90Y) used in targeted radionuclide therapy is usually imaged with SPECT by recording the bremsstrahlung X-rays following the beta minus decay. As an alternative, 90Y can be imaged with PET by detecting annihilation photons generated after internal pair creation. The e- e+ pair is produced in 0,0032% of the decays [1]. The aim of this study was to present the possibility of PET imaging of 90Y and to compare 90Y SPECT and PET images. Materials and methods: The IEC Body Phantom with 90Y in 6 spheres was used. The activity in the spheres was 3,8 MBq/ml while in the background 0,038 MBq/ml. A GE Discovery ST scanner was used for PET/CT imaging. To prevent saturation of the detectors, a copper ring was inserted into the PET gantry to absorb the bremsstrahlung photons. Thickness of the ring was 2 mm. Time of a single PET scan was 30 min. For SPECT/CT imaging a GE Infinia Hawkeye 4 with HEGP collimators was used. The energy window was 40-240 keV. Time of SPECT acquisition was 15 min. Sensitivity defined as counts/MBq s was compared by acquiring 0,2 ml radioactive source (in a syringe) at the center of the FOV. Spatial resolution was measured on a line source imaged both surrounded by air and water. In all studies matrix size was 128x128. Results: Sensitivity was about 9 times higher for SPECT than for PET. Spatial resolution was 6,1 mm for PET (values similar for air and water) versus 16,3 mm for SPECT (water) and 14,7 mm (air). The two smallest spheres (10 and 13 mm diameter) were not visible both on SPECT and PET images. Image contrast, defined as I-(counts in BCKG/counts in hot ROI), was better for PET images: 0,96 vs 0,70 for 17 mm diameter sphere, 0,98 vs 0,92 for 37 mm diameter sphere. Transverse slices of the IEC Body Phantom imaged with 90Y SPECT and PET are shown in Fig. 1.
a) I b)
ê • I m *
Figure 1. Transverse slices of the IEC Body Phantom imaged with 90Y SPECT a) and PET b). Four out of six spheres are visible on both images. Conclusion: 90Y PET/CT technique shows a better spatial resolution, but a poor sensitivity. Hence it could be use as a imaging method of choice for liver SIRT or 90Y-DOTATATE therapy, when high activity is administrated directly into a liver. Patients with lesions in the liver more than 15 mm of diameter should be studied both with PET/CT and SPECT/CT.
1. Selwyn RG, Nickles RJ, Thomadsen BR, DeWerd LA, Micka JA (2007) A new internal pair production branching ratio of 90Y: the development of a non-destructive assay for 90Yand90 Sr. ApplRadiatIsot 65:318-27
171 NUTECH-2011
SUSCEPTIBILITIES TO RADIATION OF LYMPHOCYTES FROM CANCER PATIENTS IN COMPARISON TO REFERENCE GROUPS BY CLASSIC AND MOLECULAR CYTOGENETICS
J. Miszczyk1, A. Cebulska-Wasilewska1, B. Dobrowolska2, Z. Drąg3, Z. Rudek1, W. Jędrychowski4, Z. Dobrowolski2
1 Institute of Nuclear Physics PAN, Department of Radiation and Environmental Biology, Radzikowskiego 152, Krakow, Poland 2 Jagiellonian University, Collegium Medicum, Department and Clinic of Urology, Grzegórzecka 18, Krakow, Poland 3 Jagiellonian University, Institute of Sociology, Grodzka 52, Kraków, Poland 4 Jagiellonian University, Collegium Medicum,Chair of Epidemiology and Preventive Medicine, Kopernika 7, Krakow, Poland
The genetic constitution and health status of an individual can affect his response to various exogenous (environmental, occupational, accidental or therapeutic) exposures. The aim of our study was to assess if cancer patients constitute a more radiosensitive or resistant subgroup in the population at high risk of accidental exposure (i.e. nuclear power plant failure or terroristic attack) and to examine whether this is associated to any endogenic or exogenic factors. Vulnerability to chromosome aberrations by challenging X-ray exposure detected by classic and molecular (FISH) cytogenetics was compared in lymphocytes of various cancers patients (colorectal, prostate cancer) and clinically healthy control subjects with benign prostate hyperplasia stage (BPH) or with gallbladder inflammation (GI) treated as reference groups. Results from classic cytogenetics showed a significantly lower level of chromosomal damage in BPH patients. No significant differences were observed between cytogenetic biomarkers in PCP, CCP and gallbladder inflammation subjects. These results are confirmed by a significantly higher level of chromosome 1 aberrations in PCP patients compared to BPH, detected with the FISH technique. The frequency of translocations and acentric fragments per 100 metaphases detected in chromosome 1 were also significantly higher in PC patients as compared to the BPH group. After stratification of PCP and BPH patients according to reports of cancer in the immediate family, a strong and statistically significant difference was detected between them in the frequency of deletions. Our results clearly demonstrate that the frequency of all aberrations detected in response to a challenging X-ray dose is significantly higher in subgroups of cancer patients and GI than in BPH, therefore these subgroups can be at much higher risk in case of unexpected exposure.
Acknowledgments: Research was partially supported by grant: "Studies on individual susceptibility to radiation- induced damages, DNA repair and instability of human chromosomes in lymphocytes of patients diagnosed or treated with radioiodine" (0296/B/P01/2008/3 5) and " Estimation of individual radiosensitivity of patients with prostate cancer and benign prostatic hyperplasia with application of FISH technique "(2520/B/P01/2010/39).
172 Biomedical Studies
THE SINGLE CELLS RESPONSE TO THE PROTON MICROBEAM IRRADIATION
A.Wiecheć1, J. Lekki1, E.Lipiec1, W.Polak2, W.M.Kwiatek1
lrThe Henryk Niewodniczański Institute of Nuclear Physics, Polish Academy of Sciences, Kraków, Poland, 2Medical Physics Department Royal Surrey County Hospital NHS Fundation Trust, United Kingdom
Microbeam facilities are excellent tools in the radiation biology research (particularly in cancer biology studies). They provide the possibility of targeting individual cells within a population with defined number of particles. Microbeam experiments have shown that not only the dose, but also the targeted site (i.e. nucleus or cytoplasm) determines the level of damage and cell survival. In our experiments single cells were irradiated with 2 MeV focused, horizontal, proton microbeam from the Van de Graaff accelerator. To allow cell irradiation in the atmosphere, the single ion hit facility (SIHF) was constructed. In order to determine the single cell response to the proton microbeam irradiation cells from established cell lines were used (prostate cancer PC3, normal prostate cells PZ-HPV7, gastric carcinoma MKN7, and fibroblasts ACC110). The single cells were irradiated by exact number of protons (50-8000 protons per cell) and the level of cell death induction was measured. Here we present and discuss results of following experiments: • The visualization of the necrotic and apoptotic cells (interphase death) by PI and Hoechst 33342 staining, respectively; The formation of the micronucleus (mitotic death) in gastric carcinoma MKN7 and fibroblasts ACC110 after irradiation by 500 and 3000 protons per cell; • The proton cytotoxicity in PC3 and PZ-HPV7 cells; • The observation of the morphological changes in PC3 cells after proton irradiation; • The detection of the DNA damage using fluorescent antibodies specific to double strand breaks formation (histone H2HAX) and FTIR spectroscopy. Obtained results confirmed the usefulness of the proton microbeam irradiation in cellular response studies and its importance for cancer biology research. However further research is necessary in order to translate obtained results into optimisation and individualisation of radiotherapy. The studies of the single cells response to the proton microbeam irradiation (such as kinetic of DNA damage repair, bystander effect) will be conducted in the future.
173 NUTECH-2011
THE USE OF SR-FTIR MICROSPECTROSCOPY FOR THE PRELIMINARY BIOCHEMICAL STUDY OF THE RAT HIPPOCAMPAL FORMATION TISSUE IN CASE OF PILOCARPINE INDUCED EPILEPSY AND NEUROPROTECTION WITH FK-506
J. Dudała1, K. Janeczko2, Z. Setkowicz2, D. Ei chert3, J. Chwiej1
1AGH-University of Science and Technology, Faculty of Physics and Applied Computer Science, Krakow, Poland, 2Jagiellonian University, Institute of Zoology, Department of Neuroanatomy, Krakow, Poland, 3ELETTRA, Trieste, Italy
The main aim of the work was the biochemical analysis of the hippocampal formation tissue in case of epileptic rats treated with neuroprotective agent FK-506. Three groups of animals were compared: rats with pilocarpine induced seizures treated and non-treated with tacrolimus as well as naive controls. Synchrotron radiation Fourier transform infrared (SR-FTIR) microspectroscopy was used for the analysis of the distribution of proteins, lipids as well as changes in protein secondary structure and saturation level of lipids. The measurements were carried out at SISSI beamline of ELETTRA. A Bruker IFS 66v/S interferometer coupled to Bruker Hyperion 2000 microscope was used. The tissue samples were analyzed in transmission mode with a beam defined by small aperture and spatial resolution steps of 10 |im which allowed us to probe the selected cross-line of the sample at cellular resolution. The obtained results enabled to compare the distributions of proteins and lipids in the three hippocampal cellular layers, i.e. in molecular, multiform and granular layers. For epileptic animals both treated and non-treated with FK-506 the tendency for increase of amide II/amide ratio was observed however only for multiform layer these changes were statistically significant. Similar relation was noticed in case of the ratio of the absorbance at around 1631 and 1658 cm-1. The mentioned results may suggest conformational changes of proteins in the direction of ß-sheet secondary structure. Additionally, statistically significant increase in the lipid massive and decrease of the ratio of absorbance at around 2921 and 2958 cm-1 were observed for epileptic animals treated with tacrolimus comparing to the control group.
174 NUTECH-2011
APPLICATION OF GC TO STUDY RADIOLYSIS OF CULTURAL HERITAGE ARTEFACTS
W. Głuszewski w.gluszewski@ichtj. ~waw.pl
Institute of Nuclear Chemistry and Technology, Warsaw
In response to Member State requests, the Technical Cooperation Department of the EuropeanRegion launched in 2009 the TC Project RER8015 on "Using Nuclear Techniques for the Characterization and Preservation of Cultural Heritage Artefacts in the Europe region" to improve the characterization and preservation of cultural heritage artefacts through the application of nuclear techniques with special emphasis on gamma irradiation end electron beam treatment, making use of techniques including insect eradication and disinfection in various cultural heritage materials and consolidation of degraded materials with radiation- curing resins. One of the main tasks in the radiation preservation of cultural heritage artifacts is to reduce material degradation caused by exposure to ionizing radiation. At present, disinfection procedures, and sterilization carried out using gamma rays and electron beams. One drawback of the electron beam technique is the limited scope to almost 10 cm of material with a density of 1g/cm3. EB is the advantage of high dose rate which in practice can reduce material degradation processes. Oxidation of natural polymers in atmospheric environment is responsible for degradation of properties. These processes are also initiated by increased temperature and/or by UV light. In the case of a relatively long irradiation with 60 Co sources is possible because the diffusion constant of oxygen into the material. A few seconds time to deliver the same dose of electron beam reduces this unfavorable phenomenon. Drew attention to the aromatics protection effects. In this work a comparative study using gamma rays and electron beam. Has been studied the most typical objects of historic materials: wood, paper, leather, wool and polyethylene as reference. Using gas chromatography was determined of radiation yield of hydrogen end yield of oxidation process. The gas chromatograph Shimadzu GC 2014 has been installed in air conditioned and thermostated (23.5oC) room. The column was 1 m long packed with molecular sieves 5A, the detector was thermo-conductivity (TCD - 2014) element by Shimadzu. Radiation treatment was performed at the Institute of Nuclear Chemistry and Technology in Warsaw, using cobalt gamma ray sources "Issledovatel (dose rate 0.389 Gy / h) and" GC 5000 "(8.5 Gy / h). Sample were delivered dose of 5, 10 and 25 kGy. Electron beam irradiation was performed using a linear electron accelerator, "Electronics" 10/10 an energy of 10 MeV and beam power of 10 kW. In the future we will create a summary of data on radiation resistance of materials relevant to the conservation of works of art. This directory can be developed in cooperation with the IAEA Technical Cooperation Project - RER 8015.
1. Bik J, Głuszewski W, Rzymski W.M, Zagórski Z.P, (2003) EB radiation crosslinking of elastomers, Radiation Physics and Chemistry 67, 421 2. Głuszewski W, Zagórski Z.P, Rajkiewicz M, Mikołajka A, (2010) Od Marii Skłodowskiej- Curie do radiacyjnego sieciowania elastomerów, Kauczuki w Przemyśle 1, 19-21 3. Głuszewski W, Zagórski Z.P, (2008): Radiation effects in polypropylene/polystyrene blends as the model of aromatic protection effects, Nukleonika 1, 53, 21-24
176 Nuclear Techniques in Preserving Cultural Heritage
IDENTIFICATION AND CHARACTERIZATION OF USED PIGMENTS USED IN ICON PAINTINGS FROM ST DEMETRIUS ORTHODOX CHURCH IN KORYTNIKI BY INSTRUMENTAL NEUTRON ACTIVATION ANALYSIS AND SCANNING MICROSCOPY
E. Pańczyk1, M. Pańczyk2, D. Chmielewska-Smietanko1, J. Giemza3, J. Olszewska-Świetlik4, L. Giro2
e.panczyk@ ichtj.waw.pl
'institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland 2Polish Geological Institute- National Research Institute, Rakowiecka 4, 00-975 Warsaw, Poland 3 Castle Museum in Łańcut, Zamkowa 1, 37-100 Łańcut, Poland 4The Institute for the Study, Restoration and Conservation of Cultural Heritage, Nicolaus Copernicus University in Toruń, Sienkiewicza 30/32, 87-100 Toruń , Poland
The main aim of this work is identification and characterisation of inorganic pigments from the icons of St Demetrius orthodox church in Korytniki near Przemyśl, now collected at the Castle Museum in Łańcut. The objects studied are a religious, historical and artistic phenomenon, being a part of cultural heritage of multinational Republic of Poland. These paintings of high artistic value are representative of this regional iconographic style. Samples of inorganic pigments and grounds were taken from following XVI-th century icons: The Christ Baptism, St Paraskeva Pyatnitsa, St John Evangelist and The Last Judgment. The samples were analysed by optical microscopy, SEM-EDS and INAA in order to determine the stratigraphy of art work and the identification of the pigments used. Twenty eight elements were selected for multi-parameter statistical analysis. The clustering, principal components and discriminant function analyses using STATISTICA (StatSoft) programme was carried out to identify the similarity degree of the objects analysed and the sources of the inorganic pigments. Together with commonly occurring pigments, such as vermilion (cinnabar), red lead, red iron oxide, orpiment, yellow ochre, lead white, chalk, gypsum, anhydrite and copper-containing green, one unusual material was identified - lead tin yellow. Elemental analysis, especially trace elements analysis, carried out for lead white and earth pigments, allows establishing chemical patterns or "finger prints", which are characteristic of specific artistic workshops. Extensive research aimed at determining precise details on the painting techniques applied, as well as the age, origin and authenticity of the objects examined has brought the results that could be a basis for future restoration of the paintings.
Acknowledgments: Research was supported by Ministry of Sciences and Higher Education (grant no. N N507 2066 33).
177 NUTECH-2011
INAA AND OTHER ANALYTICAL TECHNIQUES IN CULTURAL HERITAGE- ELEMENTAL ANALYSIS OF METAL THREADS FROM SILK VELVET IN WILANÓW MUSEUM-PALACE
A. Skłodowska1, H. Polkowska -Motrenko2, B. Danko2, Jakub Dudek2, Ewelina Chajduk2
:Warsaw University, Faculty of Biology, Miecznikowa 1, 02-096 Warsaw, Poland 2 Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland
Metal threads have been used in different textiles (eg. embroideries, tapestries) for thousands years but their examination with scientific methods in larger scale has started in the second half of the 20th century. The culture heritage artefacts are usually unique and only very small samples are available for analysis. Analytical techniques such as activation analysis and inductively coupled plasma mass spectrometry could be applied in research on composition and provenance of historical metal artefacts, particularly made of gold and silver. The results from the fingerprinting analysis may be helpful in determining of origin of silver and gold threads present in silk velvet (1710-1730) in Wilanów Museum-Palace. The small pieces of threads were removed from artefacts during conservation works. Instrumental neutron activation analysis was used to determine main elements in threads: gold and silver. Samples, standards and blank were irradiated for 20 min in Polish nuclear reactor MARIA (neutron flux 1014 cm-2 s-1). After 10 days of cooling, the samples and standards were unwrapped, cleaned with ethanol and subsequently the gamma-measurements were performed. The determination was carried out on the basis of activity of formed radionuclides: 110mAg and 198Au. Trace elements: Nb, Zn, Tl and Cu were determined by ICP-MS after two-steps digestion procedure. The following nuclides: 63Cu, 66Zn, 93Nb, 205Tl were selected since they are free from interference and are sufficiently abundant for quantitative measurement by ICP-MS. Knowledge about their chemical composition should help to restore original color of metal threads (flakes) and should give valuable clues about the origin of gold and silver in goldsmith workshops in Genoa (Italy) in 18th century.
178 Nuclear Techniques in Preserving Cultural Heritage
INSTRUMENTAL NEUTRON ACTIVATION ANALYSIS AS A SOURCE OF INFORMATION CONCERNING THE ORIGIN OF "RUTHENIAN ALABASTER"
T. Śliwa1, E. Pańczyk2,
[email protected]; [email protected]
1 Department of General Geology and Environment Protection, Faculty of Geology, Geophysics and Environmental Protection, AGH University of Sciences and Technology, A. Mickiewicza Av. 30., 30-059 Cracow, Poland 2 Laboratory of Materials Research, Institute of Nuclear Chemistry and Technology, 16 Dorodna St., 03-195 Warsaw, Poland
The problem of determining the source of building stone and decorative styling used in interior architecture or in the production of sculptures applies to every historical city in the world. "Each historic town has its own unique flavour. Thanks to the use of various types of stone over hundreds of years, there emerged a great stone mosaic, distinctive and unique to a given place." (Rajchel, 2007). The constant practice of import and export of finished work over the centuries increases doubts as to the attributes (Lipińska, 2007). In Poland, interdisciplinary scientific research has been undertaken on the question of materials combining petrographic methods, historical geology and economic history since the mid- twentieth century (Wadzyński, 2009). The purpose of this research was to identify historical Badenian alabaster deposits from the Ukrainian part of the Carpathian foredeep, which served as raw material for historical sepulchral and figural statues from the Renaissance era to the interwar period. Using the method of instrumental neutron activation analysis, twenty-four samples of alabaster gypsum were examined from, among others, the chapels of the Wawel Cathedral, Lvov and Tarnów, as well as from the parish churches of the Tarnów and Przemyśl diocese. Small amounts of alabaster were collected along with antique furnishings from the Roman Catholic parishes around Lvov and the Orthodox Church of St. Michael the Archangel in Lvov. Forty-five alabaster samples collected from natural outcrops and quarries that appear along the Dniester Valley in the historic regions of Eastern Galicia, Podolia and Bukovina served as reference material. Analysis of samples by INAA using the patterns of the analysed elements was carried out at the Institute of Nuclear Chemistry and Technology in Warsaw. A multi- parameter statistical analysis was performed to determine the degree of similarity between the studied objects (analysis of principal components and cluster analysis) using the STATISTICA-8 programme (Ligęza et al., 2001).
1. Ligęza M, Pańczyk E, Rowińska L, WaliśL, Nalepa B (2001) A contribution of INAA to the determination of the provenance of the fourteenth century sculpture. Nukleonika 46; 2:71-74 2. Lipińska A (2007) Wewnętrzne światło. Wydawnictwo Uniwersytetu Wrocławskiego, Wrocław 3. Rajchel J (2005) Kamienny Kraków. Spojrzenie geologa. Uczelniane Wydawnictwa Naukowo-Dydaktyczne AGH, Kraków 4. Wardzyński M (2009) Between Italy and the Low Countries. Centers of stonemasonary and sculpture in Central Europe and the early modern tradition. Some remarks on material and technology. In: Lipińska A (eds) Material of sculpture. Between technique and semantics. Wydawnictwo Uniwersytetu Wrocławskiego, Wrocław, pp 425-454
179 NUTECH-2011
PHYSICAL CHEMICAL ANALYSIS OF y-IRRADIATED WOODEN ARTEFACTS
M. Virgolici1,1.R. Stanculescu1'2, MM. Manea1, P. Bugheanu2, CD. Negut1,3, C.C. Ponta1, I.V. Moise1, M. Cutrubinis1
istanculescu@nipne. ro
1Horia Hulubei National Institute of Physics and Nuclear Engineering, IRASM Radiation Processing Centre, Physical and Chemical Tests Laboratory, 077125 Magurele, Romania 2University of Bucharest, Department of Physical Chemistry, 030018 Bucharest, Romania 3University of Bucharest, Department of Atomic and Nuclear Physics, 077125 Magurele, Romania
Gamma irradiation is an efficient method of wooden artefacts biological decontamination. Colour shifts and changes in the chemical structure induced by the irradiation treatment must be evaluated objectively and precisely. FTIR and FT-Raman spectroscopy and multivariate analysis were used to investigate the possible structural changes in experimental models of historical wooden artefacts from a 300 years old Romanian church. Thermal analysis and analytical pyrolysis were also employed as complementary tools for composition and macromolecular structure characterization [1,2]. Experimental models of 300 years old historical pinewood were irradiated at IRASM Co60 gamma irradiator with 6, 12, 25, 50 and 100 kGy. The samples were divided in three batches, two of them being subjected to artificially induced accelerated degradation in air tight inert containers (half with dry atmosphere and half with 100 % relative humidity) for 6, 12 or 18 weeks at 80 °C. Bruker Vertex 70 class FT-IR and FT-Raman spectrometer equipped with optic fibres mobile MIR probe and RAMPROBE attached to RAM II module (LN2 Ge detector, Nd:YAG laser of 1064 nm) was used. NETZSCH STA 409 PC Luxx simultaneous thermal analyzer was used for thermogravimetric analysis. Markes "UNITY" Thermal Desorber equipped with a General Purpose Hydrophobic Trap was used for direct TD of pyrolysis compounds formed at 350 °C. An Agilent GC 6890N was used with a temperature programme of 5 K/min ramp from 40 to 310 °C. MS detection was made with Agilent 5975 inert MSD in fragmentation mode by electron ionisation at 70 eV, data acquisition in SCAN mode 35 - 700 amu [3]. Small changes in the vibrational spectra of unirradiated and irradiated experimental models of historical wood samples submitted to artificial degradation were correlated with the thermal stability, the color variation and the pyrolysis compounds.
Acknowledgments: Authors are grateful to ANCS, PN II program, for financing the DELCROM project, contract no. 92-086/2008.
1. Stanculescu I, Virgolici M, Manea M, Ponta C (2008) Raman spectra of decayed wood samples, An. Univ. Buc. Chimie 17;2:41-45 2. Bratu E, Moise IV, Cutrubinis M, Negut DC, Virgolici M (2009) Archives decontamination by gamma irradiation, Nukleonika 54;2:77-84 3. Virgolici M, Ponta C, Manea M, Negut D, Cutrubinis M, Moise I, Suvaila R, Teodor E, Sarbu C, Medvedovici A (2010) TD/CGC/MS Approach for Characterization of the Volatile Fraction from Amber Specimens: a Possibility of Tracking Geological Origins J. Chrom. A 1217;12:1977-1987
180 Nuclear Techniques in Preserving Cultural Heritage
RADIATION TREATMENT OF LIBRARY AND ARCHIVAL COLLECTIONS FOR MICROBIOLOGICAL DECONTAMINATION
D. Chmielewska1, U. Gryczka1, W. Migdał1, W. Daszewski2, A. Kuberka2, Marzena Chyrczakowska2
institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland 2University of Warsaw Library, Dobra 56/66, 00-312 Warsaw, Poland
Microbiological destruction of millions volumes of books and archives has became one of the main treat in all library in the world. Moreover the microbiological burden is harmful for the librarians' and archivists' health as well. Currently the most common method used for these materials hygiene is ethylene oxide treatment, which is toxic to human and natural environment. The promising alternative for this technique can be ionizing radiation. Gamma irradiation has already been applied to the treatment of large volumes of books and archives as well as wooden sculptures. Decontamination of microbiologically infected and damaged paper can be also the area for the new application of electron beam irradiation, although this process still needs detailed investigation. In the present study influence of ionizing radiation treatment on different kind of paper was examined. Changes in chemical composition, mechanical and physical properties were studied with SEM, EDS, EPR, TGA, DSC, FT-IR methods.
181 NUTECH-2011
X-RAY FLUORESCENCE SPECTROSCOPY STUDY ON THAI DECORATIVE GLASS
K. Won-in1, S. Pongkrapan1, P. Dararutana2
[email protected] department of Earth Sciences, Faculty of Science, Kasetsart University, Bangkok 10900 Thailand 2The Royal Thai Army Chemical Department, Phaholyothin Road, Chatuchak, Bangkok 10900 Thailand
Glass has been used as decorations and ornaments in Thailand for several hundred years. In this work, the decorative glasses from historical findings as well as fragments of artifacts were characterized using X-ray fluorescence spectroscopy such as PIXE, SEM-EDX, SRXRF and |i-XRF. Results indicated that their compositions were lead-based glass. The presence of transition elements have attributed mainly to the various colorations.
182 Nuclear Techniques in Preserving Cultural Heritage
X-RAY TECHNIQUES IN THE INVESTIGATIONS OF A GOTHIC SCULPTURE 'THE RISEN CHRIST'
A. Mikołajska1, M. Walczak1, Z.Kaszowska2, M. Urbańczyk-Zawadzka3, R. P. Banyś3
Academy of Fine Arts in Cracow, Poland - Faculty of Conservation and Restoration of Works of Art, Department of Applied Physics 2Academy of Fine Arts in Cracow, Poland - Faculty of Conservation and Restoration of Works of Art, Institute of Painting Technologies and Techniques 3Center for Diagnosis Prevention and Telemedicine, John Paul II Hospital
For over a century, the X-ray radiation plays an important role in the cultural heritage area. X-ray techniques belong to the fundamental and very helpful methods used in investigation and conservation of works of art. They have been developed over the last decade thanks to the recent use of new and advanced analytical techniques. The presentation will review three different X-ray techniques applied in the investigation of a Gothic sculpture 'The Risen Christ'. Firstly, an X-ray image has been made to see the general condition of the sculpture. Thanks to the X-ray radiation properties (different absorption by various materials) we may observe the structure of an investigated object and decide which physical and chemical methods should be chosen for further investigations. An X-ray image may reveal changes or a totally different composition hidden in the bottom layers of a painting or a sculpture. We may observe changes connected with each restoration, as well as changes connected with the 'ageing' of an object. Then, for elemental analysis, a scanning electron microscopy combined with energy dispersed X-ray spectroscopy (SEM-EDX) has been used. In this technique few samples were taken from different parts of the wooden sculpture. They were embedded in epoxyde resin and polished manually using a polishing cloth of various grain-size. Using SEM-EDX, we were able to identify the painting materials (pigments and filers) in all layers of each sample. As the third and most comprehensive method, X-ray Computed Tomography (CT) has been chosen. X-ray Dual Source Computed Tomography (DSCT) is one of the most powerful non- destructive techniques, therefore it is used for the full volume inspection of an object. It has given us morphological and physical information of the inner structure of the investigated wooden sculpture.
183 NUTECH-2011
ACTIVITY OF E-BEAM IRRADIATION IN THE CONTROL OF RHIZOCTONIA SOLANI
L B. Orlikowski \ U. Gryczka 2, M. Ptaszek \ W. Migdał 2
1 Institute of Horticulture, Konstytucji 3 Maja 1/3, 96-100 Skierniewice, Poland 2 Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland
Rhizoctonia solani Kuhn is one of the most dangerous soil-borne pathogen caused pre- and postemergence damping-off seedlings and stem base and root rot of older plants. Infested soils or substrates and plant materials are the main source of that species. The aim of this study was to determine the influence of E-Beam irradiation of energy 10 MeV on in vitro and in vivo control of R. solani. Four-day-old the species cultures were irradiated with doses 0, 1.5, 3, 4.5 and 6 kGy whereas substrate from 5 to 30 kGy. Growth of cultures on treated plates and healthiness of chrysanthemum growing in irradiated substrate was estimated. Additionally, colonization of stem parts of chrysanthemum by irradiated cultures were estimated. In vitro application of ionizing radiation of 4-day-old R. solani cultures, grown on PDA medium, resulted in the lack of the species development already at dose 4.5 kGy. Inoculation of chrysanthemum stem parts with irradiated cultures resulted in the necrosis development but inocula from nontreated plates colonized plant parts significantly quicker than those treated with E-Beam. Growing of chrysanthemum cuttings in the irradiated peat resulted in the lack of stem base and leaf rot in substrate treated with doses higher than 5 kGy. On cuttings growing in composted pine bark (cpb) and its mixture with peat (1:1) disease symptoms were not observed when dose 20 kGy was applied for desinfection.
Acknowledgments: This work was financed by Ministry of Science and Higher Education (Poland) under the project NR 1205506/2009
186 Applications of Nuclear Technologies in Agriculture and Food Processing
GAMMA IRRADIATION INFLUENCE ON THE STRUCTURE OF POTATO STARCH GELS STUDIED BY SEM
K. Cieśla1, B. Sartowska1, E. Królak2, W. Głuszewski1
kciesla@orange. ichtj. ~waw.pl
institute of Nuclear Chemistry and Technology, Dorodna 16 str. 03-195 Warszawa, Poland 2Analytical Centre, Warsaw Agricultural University, Ciszewskiego 8 str., 02-786 Warszawa, Poland
Potato starch is in Poland abundant and cheap raw material widely applied in various food, pharmaceutical and technical industries. Gelling properties of starch, viscosity and stability of gels appear to be the crucial factors that affect possible application of starch in a number of industrial products. Development of the methods of starch modification and testing of the modern products that contain new hydrocolloids are still conducted. Degradation accompanied by oxidation are desirable processes that enable to obtain starches forming gels with the reduced viscosity, in relation to those formed by the native starch [1]. Such species are widely applied as components of coatings and glues in technical industries and as functional additives in the number of instant food products. Radiation modification appear to be an alternative perspective methods that might substitute chemical and enzymatic procedures, applied till now on the industrial scale. Using of ionising radiation might enable to limit the use of strong and toxic chemicals and to reduce the costs of processes in relation to the enzymatic methods. This is due to radiation induced degradation and oxidation processes leading to a significant decrease in swelling power of starch and of the resulting gels viscosity [1-3]. Accordingly, it appear interesting to carry out the studies dealing with the influence of ionising radiation also on the structure of starch gels. Solid potato starch was irradiated with a 60Co gamma rays in the gamma cell Issledovatiel using doses of 5, 10, 20 and 30 kGy. A 12-fold decrease in viscosity was found already after irradiation with a dose of 10 kGy. Swelling power decreases linearly with an radiation dose applied and at 30 kGy reach ca 0.35 % of that determined before irradiation. Gels containing 10% of dry matter were prepared on the way of heating the starch suspensions for 45 min in the heating chamber stabilised at 100 oC. Four procedures were applied for preparation of the gels in regard to SEM studies. The DSM 942 Scanning Electron Microscope (Zeiss-Leo production, in INCT) and the Quanta 200 Microscope (FEI, in AC WAU) were used for SEM observations of surfaces and fractures of the samples. Differences were observed between SEM images recorded for the non-irradiated gels and those irradiated [4]. Images of the non-irradiated gels indicate generally a honey-comb structure. Smooth areas but with oriented fractures has appeared after irradiation. Modification of gel structure corresponds to the applied dose. The differences in structural properties of gels shown by SEM result probably due to the radiation induced weakening of the internal forces in gels and increase in their homogeneity.
1. Cieśla K (2009) Transformation of supramolecular structure initialised in natural polymers by gamma irradiation. Institute of Nuclear Chemistry and Technology, Warszawa. 2. Cieśla K (2003) Gamma irradiation influence on wheat and flour gelatinisation studied by DSC and amylography. J. Thermal Anal. Cal.,74:1271-1286 3. Cieśla K, Elias son A-C (2007) DSC studies of gamma irradiation effect on the amylose-lipid complex formed in wheat and potato starches. Acta Alimentaria, 36;1:111-126. 4. Cieśla K, Sartowska B, Królak E, Głuszewski W (2007) Gamma irradiation influence on structure of potato starch gels studied by SEM Annual Report of the Institute of Nuclear Chemistry and Technology 2006, 49 - 52 187 NUTECH-2011
ISOTOPIC METHODS FOR FOOD AND BAVERAGE AUTHENTICITY CONTROL - TRENDS AND STANDARYZATION
R. Wierzchnicki
Stable Isotope Laboratory Institute of Nuclear Chemistry and Technology, Warsaw, Poland
Nowadays in Europe, origin and authenticity is probably one of the most important characteristic features of food. This is connected with economic politics of EC in agriculture sector. Adulteration as an addition of the artificial (prohibited) components to natural products, a cheaper product put into a place expensive one and mislabeling is still frequently meet fraud. One of first, the SNIF-NMR method applied to the detection of chaptalization of wine with beet sugar was accepted by the European Community as an official method of wine analysis (EC regulation 2676/90) - see table.
Table. The isotopic methods accepted as standards for authenticity control.
Product Isotopic Parameter Method Reference
Ethanol, (D/H)Ia), Wine SNIF-NMR EC Regulation 2676/90, annex 8 (D/H)IIb), Rc) Wine Ethanol, 13C IRMS EC Regulation 440/2003, annex 2 OIV Resolution ENO 2/96 Wine Water 18O IRMS (Regulation EC 822/97) Ethanol, (D/H)I, Fruit juice SNIF-NMR AOAC method 995.17 (D/H)II, R Fruit juice Ethanol, 13C IRMS AOAC method 2004.01 Fruit juice Sugar, 13C IRMS ENV 12140 (CEN/TC174N108)
Fruit juice Water 18O IRMS ENV 12141 (CEN/TC174N109) Honey and proteins, Honey IRMS AOAC method 991.41 13C IRMS -Isotope Ratio Mass Spectrometry; SNIF-NMR Site-Specific Natural Isotope Fractionation determined by Nuclear Magnetic Resonanse
Important limitation of the application isotopic method for food authenticity control is a lack of database of stable isotope composition in different origin food. Stable Isotope Laboratory of INCT from many years carry out a study of isotopic composition of food for elaboration and implementation new IRMS methods and database for some food from polish market. The results and current trends in isotopic methods in food authenticity control will be presented.
188 Applications of Nuclear Technologies in Agriculture and Food Processing
STUDY OF STABLE ISOTOPE COMPOSITION OF CHOSEN FOODSTUFFS FROM THE POLISH MARKET
K. Malec-Czechowska, R. Wierzchnicki
Stable Isotope Laboratory, Institute of Nuclear Chemistry and TechnologyWarsaw, Poland
The isotope ratio mass spectrometry ( IRMS ) methods play a very important role in food authenticity and origin control. Stable isotope analysis for the control of declared origin is already routinely applied in EU for wine, fruit juice and honey. For other foodstuffs, published studies already exist and demonstrate the potential for verification of origin and authenticity for milk and milk products, olive oil, asparagus and meat ( pork, beef, lamb ). In the present work are demonstrated the carbon and nitrogen isotope ratio in samples of food products bought in the retail trade in Warsaw. Research were carried out on meat ( pork, chicken ), hens eggs and honey. These products were originated from the conventional farming and from ecological farms. The meat and eggs samples were dried by lyophilisation and the fat was removed in a Soxhlet apparatus using petroleum ether. The remaining fat-free material (protein) was homogenised. Proteins from honey were isolated according to the AO AC Official Method 998.12 and own procedure. C and N isotope analysis (d 13C and d 15N ) were performed on DELTA plus (Finnigan, Germany) mass spectrometer using elemental analyser Flash 1112 series EA. The values of the isotopic ratios are expressed in d%o and correspond to an international standard ( V-PDB for S13C, and Air for d 15N ) according to the following general formula: d%o = R sample - R standard / R standard x 1000 where R represents the ratio between the less abundant and more abundant isotopes, in particular 13C/12C and 15N/14N. The results received in our laboratory were compared with results presented in the literature for suchlike products. The results show that the stable isotope ratios of bio-elements in tested products can be applied to verify the origin of products.
189 NUTECH-2011
STUDY ON RADIATION INDUCED RADICALS GIVING RISE TO EPR SIGNALS EMPLOYED FOR THE DETECTION OF RADIATION TREATMENT IN SUGAR CONTAINING FOOD
G.P.Guzik, W.Stachowicz, J.Michalik
Institute of Nuclear Chemistry and Technology, Warsaw, Poland
The present investigations deal with simple C6H12O6 sugars, D-mannose and L-sorbose separated from cranberry (Oxycoccus) and rowan-berries (Sorbus aucuparia) and exposed to ionizing radiation. Irradiation is applied by some food exporters in Asia and Africa for the conservation and disinfection of fruits both fresh and candied. The method is recommended by FAO/WHO Codex Alimentarius Vol. XV while irradiated fruits have to be labelled. However, this treatment is not permitted in EU. Irradiation produces in food containing crystalline sugars (candied fruits for example) stable but still not identified radicals giving rise in EPR to a complex but specific signals. Currently these signals are brought in the practice for detection of fruits treated by radiation. The aim of present study is the identification of radicals produced by radiation in fruit sugars. Both sugars were irradiated in Co-60 source with the dose of 4 kGy of gamma rays. Thereafter, the EPR signals of samples stored at 20oC were recorded. Subsequently, the EPR signals of parallel samples heated at the temperatures below but close to the melting point of investigated sugars (mannose 95oC, sorbose 140oC) were registered too. Computer controlled subtraction of normalized EPR spectra both heated and unheated results in differential spectrum corresponding to less stable fraction of radicals. Basing on spectral parameters derived from differential spectra the DFT (density functional theory) calculations were accomplished to assign the experimental spectra to adequate radical structure. The EPR spectra recorded with D-mannose for both heated and unheated samples did not show any significant differences. The basic pattern remains unchanged after heating while intensity is significantly suppressed. It is concluded, therefore, that all radicals responsible for a complex EPR spectrum of irradiated D-mannose are of the same thermal stability. The EPR signal of D-mannose is relatively simple as compared with the spectra of the other sugars. A dominating signal of the spectrum is a doublet with A iso « 3,0 mT. The DFT calculations show that this doublet represents presumably the radical with unpaired electron at C-3 interacting with nuclear spin of H atom bound to C-2 (calculated A iso = 2,69 mT). The heating of L-sorbose results in the decay of central lines of the signal while the outer lines from both size of the signal remain practically unchanged. The resultant differential spectrum of L-sorbose shows a dominant doublet with A ¡so « 2,6mT. The DFT calculation suggests that such doublet could be assigned to a radical with unpaired electron at C-4 interacting with nuclear spin of H atom bound to with C-3. The present study based on our original method combining high temperature treatment with theoretical DFT calculation brought new data on the composition of stable radicals produced by radiation in sugars.
190 NUTECH-2011
A COMPARATIVE STUDY ON THE PERFORMANCE OF RADIATION DETECTORS FROM THE HgI2 CRYSTALS GROWN BY DIFFERENT TECHNIQUES
J.F.T. Martins, F. E. Costa, R.A. Santos, C. H. Mesquita and M. M. Hamada
Energy and Nuclear Research Institute, CNEN/SP, São Paulo, Brazil
There have been attempts to develop room-temperature X- and gamma ray semiconductor detectors for various applications. The main physical semiconductor properties required for fabrication of room temperature semiconductor detectors are: (1) high atomic number; (2) high density; (3) high absorption coefficient; (4) a band gap large enough to keep leakage currents low, at room temperature and (5) large electron and hole mobility-lifetime products, for an efficient charge collection [1, 2]. Among these types of detectors, HgI2 has emerged as a particularly interesting material in view of its wide band gap (2.13 eV) and its large density 3 (7.5 g/cm ). HgI2 crystals are composed of high atomic number elements (ZHg=80 and Zi=53) and with high resistivity (>1014 ficm). These are important factors in applications where compact and small thickness detectors are necessary for X- and gamma rays measurements. However, the applications of Hgi2 are limited by the difficulty in obtaining high-quality single crystals and the long-term reliability problems in devices made from crystals [1]. in this work, the Hgi2 crystals were grown using four different techniques: (a) physical vapor transport, (b) solution from dimethyl sulfoxide complexes, (c) vapor growth of HgI2 precipitated from acetone and (d) Bridgman method. The obtained crystals for four methods were characterized considering the following physical chemistry properties: crystal stoichiometry, crystal structure, plan of the crystal orientation, surface morphology of the crystal and crystal impurity. The influence of these physical chemistry properties on the crystals developed by four techniques was studied, evaluating their performance as a radiation detector. The best result of radiation response was found for the crystal grown by physical vapor transport. Also, the dependence of the radiation response on the HgI2 crystal purity was also studied. For this, the HgI2 raw material was purified by the many pass zone refining technique. A significant improvement in the characteristics of the detector-crystal was achieved, when the starting materials became purer.
1. McGregor DS, Hermon H (1997) Room-temperature compound semiconductor radiation detectors. Nucl. Instr andMeth. Phys. Res. A 395: 101-124 2. Oliveira IB, Costa FE, Chubaci JFD, Hamada MM (2004) Purification and preparation of TlBr crystals for room temperature radiation detector applications. IEEE. Trans. Nucl. Sci 51: 1224-1228
192 Technological Developments
ACCREDITED LABORATORY FOR MEASUREMENTS OF TECHNOLOGICAL DOSES (LMTD)
A. Korzeniowska - Sobczuk, K. Doner and M. Karlińska
Institute of Nuclear Chemistry and Technology, Dorodna 16, 03-195 Warsaw, Poland
Laboratory for Measurements of Technological Doses is an organisational unit of the Institute of Nuclear Chemistry and Technology. This accredited laboratory specializes in development and implementation of the latest methods for technological dosimetry of ionizing radiation. Since 2004 LMTD has the accreditation of Polish Centre for Accreditation (AB 461), also has implemented a quality management system meeting the requirements of "General requirements for the competence of testing and calibration laboratories (ISO/IEC 17025:2005)". Scope of the accreditation includes: measurements of high-doses (from Gy to kGy) of gamma radiation, X-ray and high-energy electron beam; irradiation of dosimeters, and other small-size samples with well-defined doses of gamma radiation from 60Co-source and electrons of energy 10 MeV. LMTD provides a high level and quality of research, create and develop new methods for measuring high doses of ionizing radiation and provides services, among others the irradiation of small-sized samples with well-defined doses of gamma radiation or with high energy electrons and measurement of electron energy and dose distribution for deep-study research. Activities of LMTD are used in many fields such as medicine (radiation sterilization of medical devices and allografts), ecology (biological materials), plastic industry (radiation modification of polymeric materials). Currently, the following activities are conducted performed: - traceability of dose measurements for routine dosimetry in radiation sterilization of medical devices using graphite and polystyrene calorimeters and the sensitivity investigation of dosimeters used for routine film dosimetry process (ISO 11137-3 standard); - development of the methods for controlling the sterilization of tissue allografts [1], [2] carried out at a temperature of dry ice using film dosimeters B3, PCV, CT A, calorimeters - in ambient temperature; - dosimetry of sealed 60Co sources (using Fricke dosimeter and CTA) for later use in: radiation modification of polymers, radiation biology and medicine; - estimation the dose distribution homogeneity in the products of biological importance; - dose mapping performed during Operational and Performance Qualification. (ISO 11137-3 standard).
1. Z. Pelmel-Stuglik; Paragraph 14 in „ Validation of radiation sterilization process of tissue allografts", Warsaw 2009, Project financed by EU 2. Z. Pelmel-Stuglik, S. Fabisiak, R. Adamska, M. Karlińska; Paragraph 13 in „Radiation sterilization process of tissue allografts" Warsaw 2009, Project financed by EU
193 NUTECH-2011
COMPARISON OF GEANT4 SIMULATIONS WITH EXPERIMENTAL PROTON ENERGY LOSS FOR SOME THICK ABSORBERS
1 1 1 2 3 2 2 O. Yevseyeva , J.T. de Assis , I. Ievsieieva , H.R. Schelin ' , I. Evseev , E. Milhoretto , F. da Silva Ahmann2, S.A. Paschuk2, V. V. Denyak2'3, K. S. Diaz4, J.M. Hormaza5, R.T. Lopes6 [email protected]
instituto Politécnico da Universidade do Estado do Rio de Janeiro - IPRJ, Brazil 2Federal University of Technology-Paraná - UTFPR, Curitiba - PR, Brazil 3Pelé Pequeno Príncipe Research Institute, - IPP, Curitiba-PR, Brazil 4Centro de Aplicaciones Tecnologicas y Desarrollo Nuclear, Havana, Cuba 5Instituto de Biociências de Botucatu da UNESP, Botucatu - SP, Brazil "^Laboratório de Instrumentação Nuclear, COPPE, UFRJ, Rio de Janeiro - RJ, Brazil
Although the physics of proton interaction with matter for thick absorbers has a well- established theory for the so-called Bethe-Bloch domain [1], the basic principles of Monte Carlo simulation for such processes are well known since the middle of the past century [2], and GEANT4 has been validated against proton stopping powers from the NIST PSTAR [3], the evolution of the code leads to some result instability within the various code releases. In this work, we present the recent results for the comparison of our GEANT4 simulations against experimental proton energy loss for some thick absorbers. All the simulations were performed using the GEANT4 Hadrontherapy Advanced Example. The GEANT4 versions 4.8.2, 4.9.2, and 4.9.4 were tested with different simulation parameters, such as varied cut values. In addition to the Standard model, some other models for the electromagnetic processes from the GEANT4 Low Energy Extension Pack were tested as well. Experimental data were taken from [4] for polyethylene, and from [5] for aluminum and gold absorbers. The theoretical predictions for the spectra were calculated using the self-consistent Gaussian solution of the Boltzmann kinetic equation in the Fokker-Plank form [1]. In order to compare the GEANT4 simulations with other popular codes, the same spectra were simulated by TRIM/SRIM2011 and MCNPX2.4.0 [6]. The simultaneous comparison of the results obtained for different materials at various initial proton energies were done using the reduced calibration curve approach [7].
4. Remizovich VS, Rogozkin DB, Ryazanov MI (1986) Analytic description of the penetration offast charged particles in matter. Sov. J. Part. Nucl. 17;5:409-432 5. Ziegler JF, Biersack JP, Ziegler MD (2008) SRIM: The Stopping and Range of Ions in Matter. SRIM Co., Chester-MD, USA 6. Amako K, Guatelli S, Ivanchenko VN, et al. (2005) Comparison of Geant4 Electromagnetic Physics Models Against the NIST Reference Data. IEEE Trans. Nucl. Sci. 52;4:910-918 7. Ito A, Koyama-Ito H (1984) Possible use of proton CT as a means of density normalization in the PIXE semi-microprobe analysis. Nucl. Instr. andMeth. B3:584-588 8. Tschalär C, Maccabee HD (1970) Energy-straggling measurements of heavy charged particles in thick absorbers. Phys. Rev. B1:2863-2869 9. MCNPX User's Manual (Version 2.4.0), Los Alamos National Laboratory, LA-CP-02- 408, USA, 2002 10. Yevseyeva O, Assis JT, Evseev IG, et al. (2010) Comparison of proton energy loss in thick absorbers in terms of a reduced calibration curve. Doi:10.1016/j.nima.2010.08.083
194 Technological Developments
COMPUTER SIMULATIONS AND IMAGE RECONSTRUCTION FOR A PROTON COMPUTED TOMOGRAPHY SYSTEM
E. Milhoretto1, H. Schelin1,2, J. Setti1, V. Denyak1,2, S. Paschuk1, I. Evseev1, F. Silva1, J. de Assis3, O. Yevseyeva3, R. Lopes4 and U. Vinagre Filho5
[email protected], [email protected]
1 Federal University of Technology-Paraná - UTFPR, Curitiba, PR, Brazil 2 Pelé Pequeno Príncipe Research Institute - IPP, Curitiba, PR, Brazil 3 Instituto Politécnico da UERJ, Nova Friburgo, RJ, Brazil 4 Laboratório de Instrumentação Nuclear, COPPE/UFRJ, Rio de Janeiro, RJ, Brazil 5 Instituto de Engenharia Nuclear - IEN/CNEN, Rio de Janeiro, RJ, Brazil
This work presents the results of computer simulations for proton computed tomography (pCT) for a prototype installed at the Loma Linda University Medical Center, USA, that works with proton beam with energy ranging from 200 to 250 MeV [1]. Monte Carlo simulations with codes SRIM and Geant4 are a powerful tool to estimate the proton energy loss and the straggling in typical objects used in medical applications [2][3]. For the high energy region of protons only a few experimental data are available for the pCT case. The simulations were done using the code Geant4 based on a real phantom constituted of an acrylic tube with about 20 cm diameter, a little polyethylene bar located at the center, and water [4]. The polyethylene bar can be changed for different sizes and materials. The simulations were performed in order to adjust the parameters of a previous simulation used for an experiment done at the proton accelerator at IEN/CNEN and to understand some specific physics effects that affect the form of the final proton low energy spectra [4][5]. CT images were reconstructed with Fast Fourier Transform (FFT) algorithms using the simulated data. C++ Builder was used for the algorithm that makes the conversion of equivalent thickness of water density to reconstruct the image.
1. R. W. Schlüte, et al. (2004) "Conceptual Design of a Proton Computed Tomography System for Applications in Proton Radiation Therapy", IEEE Trans. Nuc. Sei., Vol.51, p.866, 2. I. Evseev, T. Assis, O. Yevseyeva, et al., (2007) "Comparison of some popular Monte Carlo Solutions for proton transportation within pCT problem ", In: Proceedings of the 2007 International Nuclear Atlantic Conference - INAC2007, Santos-SP, Brazil. ABEN, 2007, E05-1826. 3. Yevseyeva, Olga ; ASSIS, et al., Comparison of geant4 simulations with experimental data for thick al absorbers. AIP Conference Proceedings, v. 1139, p. 97-101, 2009. 4. Milhoretto, Edney ; Schelin, H. R, et al., (2010) Geant4 simulations for low energy proton computerized tomography. Applied Radiation and Isotopes, v. 68, p. 951-953. 5. Setti, J. (2007) "Preliminary Residís of the pCT Scanner Testing aí CV-28. INAC 2007 Conference, Santos-SP, Brazil. 6. I. Evseev, T. Assis, O. Yevseyeva, et al., (2005) 'Proion CT setup at CV-28 of IEN/CNEN", Brazilian Journal of Physics, vol. 35, No.3B, pp. 747-750.
195 NUTECH-2011
DIAMOND DETECTOR FOR A SPECTROMETRIC MEASUREMENT OF "LOST ALPHA PARTICLES"
J. Dankowski1, K. Drozdowicz1, B. Gabańska1, A. Igielski1, A. Kurowski1, B. Marczewska1, T. Nowak1, U. Woźnicka1
lrThe Henryk Niewodniczański Institute of Nuclear Physics Polish Academy of Sciences, Radzikowskiego 152, PL-31-342 Kraków, Poland
A synthetic high purity CVD (Chemical Vapour Deposition) monocrystalline diamond detector has been tested in aspect of measurement of the alpha particles escaping from the deuterium-tritium hot plasma in tokamak or stellarator. This research confirms potential application of such detectors for spectrometric diagnostic of the lost alpha particles in future thermonuclear power plant reactors where this detection is considerable to get important information on the energetic balance from the thermonuclear D-T reaction in burning plasma [1]. The alpha particles brake away from plasma and can take part of energy outside and cool it [2]. The harsh environment near the first wall of tokamak, like high fluxes of particles and high temperature, make diamond an ideal candidate as a fast ion detector, especially for application in these hard conditions. Moreover the diamond crystal has interesting properties like low atomic number, large band gap, large saturated carrier velocity, low noise and radiation hardness. Spectrometric properties of the synthetic high purity CVD monocrystalline diamond detectors of the 50 |j,m thickness and active area diameter of the 2 mm have been tested using the triple alpha isotopic source AMR33 (239Pu,241Am, 244Cm, energies at -5.5 MeV). In this stage we have obtained very good results comparing to a classical silicon detector. In the next step the diamond detectors have been investigated using a Van de Graaff accelerator. In this measurement we used the RBS geometry (Rutherford Backscattering Spectrometry) to get better energy resolution and dodge the issue of high flux of particles. Accelerated helium ions have been backscattered on the thin Au foil. The alpha particles have been measured at a well defined angle with monoenergetic ion beam chargeable between 0.5 and 2 MeV. All the points lie almost perfectly on the calibration line which has been obtained from measurements with the monoenergetic beams only. In the same geometry we investigate diamond detector response for proton beam. We also test a behavior of the diamond detector in a mixed radiation filed in a 14 MeV neutron generator which acts in this case mainly as a source of alpha particles from the D-T reaction. Other reactions in the target also occur and we measure energetic spectra from three main nuclear reactions: 3H(d,n)4He, 2H(d,p)3H and 3He(d,p)4He in high flux of deuterons beam. The results are very promising in aspect of application of diamond detectors for spectrometric diagnostic of the lost alphas which maximum energy from the thermonuclear reaction is equal to 3.5 MeV.
1. Aymar R (2002) ITER R&D: Executive Summary: Design Overview, Fusion Eng. Design 55; 107-118 2. Llewellyn Smith C. (2009) The path to fusion power, Eur. Phys. J., Special Topic 176; 167-178
196 Technological Developments
EVALUATING THE SHIELDING PARAMETERS FOR NEUTRON FLUENCE OF 252CF SOURCE USING MCNP4C CODE
M.N. Nasrabadi
Department of Nuclear Engineering, Faculty of Advanced Sciences & Technologies, University of Isfahan, Isfahan 81746-73441, Iran
In this work, neutron fluence and equivalent dose rate have been calculated for 252Cf source in the energy range from 10-14 to 16 MeV in water, paraffin, polyethylene, concrete, graphite, poly-boron, borated polyethylene and boron carbide, using MCNP code. Since the MCNP code considers all kind of interactions of neutron with matter, neutron fluence and equivalent dose rate calculated and presented in this work are accurate enough to be used in neutron shielding calculations. The neutron equivalent dose rate calculated here has also been compared with experimental measurements and showed good agreement. With analysis of the MCNP code results, it was concluded that amongst the tested shields boron carbide proved to be the best shield against the neutron fluence of the 252Cf source.
1. Karelin YA, Gordeev YN, Karasev VI, (1997) Appl. Radiat. Isot. 48:10-12; 1563-1566 2. Chilton AB, Shultis JK, F aw RE, (1984) Principles of Radiation Shielding, Prentice- Hall, INC., Englewoodcliffs 3. Briesmeister JF, (2000) MCNP-A Monte Carlo N-Particle Transport Code, Version 4c ", Los Alamos NATIONAL Laboratory, Report LA-13709-M 4. Gehrke RJ, AryaeinejadR, (2004) Nucl. Instr. AndMeth. B 213; 10-21 5. Da Silva AX, V.R. Crispim, (2001) Applied Radiation and Isotopes ; 217-225 6. Mollah AS, Ahmad GU, (1992) Institute of Nuclear Science and Technology, Atomic Energy Research Establishment 7. Carilo HRV, Gallego E, (2006) Nuclear Studies and Electric Engineering Academic Units. Second European IRPA Congress, Paris 15-19 May
197 NUTECH-2011
GAMMA AND NEUTRON MAZE EFFICIENCY ENHANCEMENT
Omi, N. M., Rodrigues Jr., O., Calvo, W. A. P., Rela, P.R.
Instituto de Pesquisas Energéticas e Nucleares (IPEN-CNEN/SP),Av. Prof. Lineu Prestes, 2242 05588-900, São Paulo, SP, Brazil
The design of radiation bunkers is based on the existence of thick walls and mazes, ducts with bends, in penetration points. These mazes tend to have traditional designs with three or more bends with plain finished walls, penetrating the bunker walls with low radiation transmission rate. The traditional designs require wide installation footprints and long transport systems, following the respective maze. Reducing the radiation transmission of each bend, it's possible to reduce the amount of curves and the length of their ducts, or legs, to reach the desired maze radiation transmission rate. Minimizing the maze legs can reduce significantly the costs of the installation construction and maintenance. The proposed maze modifications, effective for gamma and neutron fluxes, have the radiation transmission many times lower than a traditional maze with the same duct number and lengths. The main purpose of this work is to show that it is possible to enhance the maze efficiency to reduce the radiation transmission adopting radiation "traps". Each "trap" have to be optimized for the radiation source and maze section. As example, the efficiency enhancement of one of these modifications for gamma radiation is determined by comparative simulations. The results of Monte Carlo simulations are presented in this work.
Keywords: gamma, neutron, maze, efficiency, enhancement
198 Technological Developments
GASEOUS DETECTORS IN CURRENT HIGH ENERGY PHISICS EXPERIMENTS
S. Koperny, T. Z. Kowalski
AGH University of Science and Technology, Faculty of Physics and Applied Computer Science, Kraków, Poland
Gaseous detectors are still in wide usage in high energy physics experiments, especially in subsystems with very large surface for example muon detection systems. Their energy resolution is just reasonable, FWHM ~ 16% for 5.9 keV. Detection efficiency for minimum ionizing particles is higher than 97%. Position resolution can be reached as low as ~ 30pm. Charge collection time is a critical parameter for detectors working in the LHC environment and for properly selected mixtures is about 50 ns, it means that these detectors can work in very harsh radiation environments up to 20 MHz. For some of them the pulse rise time is ~ 2 ns and time of discharge is ~ 10 ns, so they can be used for triggering. High modularity of detectors and wide spectrum of their shapes enable wide range of their application, they are used for continuous tracking of particles, energy reconstruction and particle identification. The following types of detectors will be presented: • Monitored Drift Tube, • Cathode Strip Chamber, • Thin Gap Chamber, • Resistive Plate Chamber, • Time Projection Chamber, CMOS Pixel Gas Detector. Their advantages and disadvantages, operating principles and construction will be summarized.
199 NUTECH-2011
INCRUSTATION OF a-PARTICLE EMITTERS IN THE SOURCE BACKING: INFLUENCE ON ACTIVITY MEASUREMENTS
A. Fernández Timón1, M. Jurado Vargas2
^SCET, Universidad Rey Juan Carlos, C/Tulipán s/n, 28933, Móstoles, Madrid, Spam 2Departamento de Física, Universidad de Extremadura, Avda. Elvas s/n, 06071 Badajoz, Spain
Alpha particles emitted from radioactive sources are often measured using a 2n counting geometry, in order to determine the activity with a low uncertainty. However, the ratio C27/N0 (counting rate/activity) can deviate greatly from the theoretical value of 0.5, because some of the alpha-particles initially emitted downward towards the backing material can be backscattered into the 2TZ detector, while other particles emitted upward towards the detector are scattered and/or absorbed in the source. For this reason, some corrections to the experimental counting rate must be performed in order to determine the real source activity. These corrections have been studied for sources of negligible and of non-zero thickness by several workers [1-3], taking into consideration Gaussian scattering models or by using Monte Carlo procedures to simulate the interaction of alpha particles in the source. However, the corrections needed for the situations corresponding to alpha-particle sources where the radionuclides are not deposited, but incrusted in the backing material have not been considered. The aim of this work is then to study the influence that this incrustation in the source can have on the total detection efficiency and, as a consequence, on the activity estimated for the source. To carry out this study, we used a Monte Carlo simulation procedure to model the behaviour of the alpha particles in the source. In particular, the well established Monte Carlo computer code SRIM [4], developed to simulate the transport of ions in matter, was applied to the study of these corrections in those cases where alpha-particle emitters are incrusted in the source backing. The values of the total detection efficiency C2/N0 were calculated for sources with three backing materials with very different atomic numbers: Al, Ag, and Pt. For each backing, several alpha-particle energies were considered, as well as a wide range of incrustation depths in the source support. The deviations from the real value of the activity if the incrustation of nuclides is not considered are also estimated.
1. Crawford J. A (1949) Theoretical calculations concerning backscattering of alpha particles, in: The Transuranium Elements, Part II (McGraw-Hill, New York). 2. Lucas L, Hutchinson J.M.R (1976) Study of the scattering correction for thick uranium- oxide and other alpha-particle sources-1: Theoretical. Appl. Radiat. Isot. 27:35-42 3. Jurado Vargas M, Fernández Timón F (2004) Scattering and self-absorption corrections in the measurement of a-particle emitters in 2 n geometry. Nucl. Instrum. Methods. Phys. Res B 217:564-571 4. Ziegler J.F, Ziegler M.D, Biersack J.P (2010) SRIM - The stopping and range of ions in matter. Nucl. Instrum. Methods Phys. Res. B 268:1818-1823
200 Technological Developments
INFLUENCE OF GROWTH PARAMETERS ON TERMOLUMINESCENT PROPERTIES OF CVD DIAMOND
M. Mitura-Nowak, A. Karczmarska, B. Marczewska, M. Perzanowski, and M. Marszałek
The Henryk Niewodniczański Institute of Nuclear Physics, Polish Academy of Sciences, Krakow, Poland
Diamond is believed to be an attractive material for thermoluminescence TL dosimetry applications due to its outstanding properties: tissue equivalence, mechanical and chemical resistance, radiation hardness [1,2]. Growth conditions strongly affect morphology and structure of polycrystalline diamond and its thermoluminescent properties [3,4]. Effects of the deposition parameters on grain size, growth rate and quality of CVD diamond were investigated. Diamond films of 2 inch diameter were deposited using a microwave plasma enhanced chemical vapor deposition (MPECVD) method with methane-hydrogen-oxygen gas mixture on silicon (100) wafers. The pressure of working gases was changed in the range of 95-125 Torr and the oxygen concentration varied from 0 to 3%. The surface morphology and structure of diamond layers (thickness about 20 pm) were investigated by Scanning Electron Microscopy (SEM), Atomic Force Microscopy (AFM), Raman spectroscopy and X-ray diffraction. In order to investigate basic TL properties, samples were exposed to gamma radiation with a doses in the range of 1 to 90 Gy. The glow curves of irradiated CVD diamond displayed a maximum of amplitude centered at 240oC. Two-dimensional (2-D) dose distribution was measured using 2-D TL reader equipped with a sensitive 640x480 pixel charge coupled device (CCD) camera. TL properties of diamond detectors, such as sensitivity, linearity, repeatability and uniformity of the signal on the sample area, were investigated. In this work, good linearity of the detector response, which is the important feature of the detectors used for dosimetry, and the high sensitivity of the signal were observed. However, the appearance of inhomogeneous distribution of the CVD diamond layers related to the growth condition parameters requires further studies.
1. Borchi E, Furetta C, Kitis G, Leroy C, Sussmann R S, Whitehead A J (1996) Assessment of Diamond as a Thermoluminescence Dosemeter Material. Radiation Protection Dosimetry 65;1-4:291-295 2. Benabdesselam M, Iacconi P, Briand D, Butler J E (2000) Performance of CVD diamond as a thermoluminescent dosemeter. Diamond and Related Materials 9; 1013 101 3. S.W.S. Mc Keever, Thermoluminescence of Solids, Cambridge University Press, 1985. 4. Gastélum S, Cruz-Zaragoza E, Chernov V, Meléndrez R, Pedroza-Montero M, Barboza- Flores M On the use of MWCVD diamond as thermoluminescent gamma dosimeter. Nuclear Instruments and Methods B 260;2: 592-59
201 NUTECH-2011
MATERIALS FOR RADIATION DETECTORS AND DOSIMETERS - THEMOLUMINESCENCE PROPERTIES OF RARE EARTH DOPED SCINTILLATING CRYSTALS
P. Bilski1, A. Twardak1 and Y. Zorenko2
1-Institute of Nuclear Physics Polish Academy of Science, Kraków, Poland 2- Institute of Physics Kazimierz Wielki University in Bydgoszcz, Bydgoszcz, Poland
One of the most widely exploited methods for radiation detection and dosimetry are those based on luminescence phenomena. Among them the most important for practical applications are scintillation and thermoluminescence. Very often both these phenomena may be observed in the same materials. Scintillating materials require short decay time of luminescence. Presence of defects in the crystal lattice is unwanted, as they may act as metastable trapping centres at room temperature range and induce a delay in delivery of charge carriers to emission centers. Sometimes such delay is very large and results even in long lasting luminescence. On the other hand, such defects are advantageous for passive dosimetry applications, where such phenomena as thermoluminescence or optically stimulated luminescence are widely exploited. Therefore, in a search for new scintillating materials, thermoluminescence is a useful tool for understanding charge trapping processes in these materials. In the same time, if deep stable trapping centres are present in the studied material that may be a special goal of research to use such centers for thermoluminescent dosimetry. Rare earth doped wide band gap oxide crystals are attractive materials for scintillating applications. In 3 the present work several samples of Ce + doped Y3Al5O12 yttrium-aluminium garnet (YAG), Lu3Al5O12 lutetium-aluminium garnet (LuAG) and LuSiO5 lutetium ortho-silicate (LSO) were studied. The samples were prepared with various techniques: Czochralski method for growing of single crystal, liquid phase epitaxy for crystallization of single crystalline film and high-temperature vacuum sintering for optical ceramic preparation. Such different methods of material preparation resulted in the presence or absence of main host defects (first of all, antisite defects and oxygen vacancies) as emission and trapping centers. The samples under study were irradiated with different doses of gamma-rays (662 keV Cs-137), alpha particles (5.4 MeV Am-241) and UV/VIS light. Thermoluminescent glow-curves were registered using the RA'94 TL reader at a linear heating ramp. The examples of the measured glow-curves are presented in the Figure 1, widely reflect the differences in the type and concentration of host trapping centers depending on material preparation.
Figure 1. Thermoluminescence glow-curves of three types of YAG:Ce samples after exposure with 50 mGy of Cs-137 gamma-rays.
202 Technological Developments
METAL-ORGANIC FRAMEWORK MATERIALS (MOF) AND THEIR APPLICATIONS
W. Starosta, B. Sartowska, A. Pawlukojć, L. Waliś, M. Buczkowski
Institute of Nuclear Chemistry and Technology, Warsaw, Poland
Intensive studies are currently carried out on metal-organic frameworks (MOF) - porous materials which are metal or metal clusters linked by multitopic organic ligands and show selective sorption properties. Interest for these materials has been driven by a possible use for greenhouse gases removal, hydrogen storage for future energy applications and gas separation. The possibility of combining a wide range of metals with similarly large number of available ligands opens ways to design the structures meeting specific purposes. Recently growing interest to nanoscale MOFs is observed. Usually high ratio surface to bulk atoms observed in the case of nanoparticles gives them novel properties. At the same time, small dimensions and possibility of functional attachment of molecules to surface or by adsorbing them inside the membrane pores opens new applications e.g. as drug carriers, imaging agents and sensors. Bulk metal-organic framework materials are synthesised by solvothermal methods. For nanoscale and thin film MOF on supporting materials, new methods have been recently reported: microfluidic synthesis, patterned growth by electrochemical synthesis, surfactance assisted growth and seeded growth. Possible applications of MOF include gas sorption and separation and application in sensor developments for detection of - for example - harmful gases. The results of research works carried out in Laboratory of Materials Research in Institute of Nuclear Chemistry and Technology on synthesis of MOF materials and their applications will be presented.
203 NUTECH-2011
MODIFICATION OF THE STRUCTURE OF THE FILMS PREPARED BASING GAMMA IRRADIATED STARCH EXAMINED BY SCANNING ELECTRON MICROSCOPY
K. Cieśla, B. Sartowska
kciesla@orange. ichtj.waw.pl
Institute of Nuclear Chemistry and Technology, Dorodna 16 str., 03-195 Warszawa, Poland
Potato starch is the appropriate material for preparation of biodegradable and edible films or coatings. Requirements of the good mechanical and barrier properties of such packaging, adequate for application in the industrial products induced necessity of improvement the films quality by using modified composition or applying various chemical and physical treatment. Our previous results show that addition of some surfactants causes increase in hydrophobicity of starch films. Moreover, improvement of the functional properties of the films prepared basing potato or wheat starches and those containing potato starch and admixed surfactants was noticed after irradiation [1,2]. In particular, improvement of hydrophobic properties of the films prepared basing potato starch, potato starch - sodium laurate, or potato starch - cetyl-trimethyammonium bromide (CTAB) composition was stated. Improvement of strength and elasticity was detected in the case of potato starch films, while in the case of the films containing addition of the surfactants the increase in elasticity was noticed. This data can be related to radiation induced starch degradation and oxidation and to modified interaction of the irradiated starch with the complexed ligands [2]. Accordingly, at present the studies of structural properties of the films prepared using potato starch alone or it's composition with both these surfactants were carried out applying scanning electron microscopy (SEM). Irradiations with Co-60 gamma radiation were carried out in the gamma cell Issledovatiel applying a dose of 30 kGy. Films were prepared on the way of casting from the gelatinised starch solutions with addition of glycerol as a plasticizer (30 wt% of glycerol in terms of starch mass). Films containing sodium laurate and CTAB were prepared after performing the procedure, leading to starch-surfactant complexes (at laurate:starch ratio equal to 0.049 and CTAB:starch ratio equal to 0.075) [2]. The results shows that the improvement of the hydrophobic and mechanical properties of the starch and starch-surfactant films resulting due to irradiation correspond to the formation of smoother and more homogeneous structure [3]. Formation of more homogeneous films can be related to formation of the smoother gels on the intermediate stage of preparation [4]. Accelerated decrease in the irradiated films elasticity after prolonged storage can be related to the facilitated evacuation of plastificator from the smaller internal spaces of those films.
1. Cieśla KA, Nowicki A, Buczkowski MJ (2010) Preliminary studies of the influence of starch irradiation on physicochemical properties of films prepared using starch and starch-surfactant systems. Nukleonika 55;2:233-242 2. Cieśla K (2009) Transformation of supramolecular structure initialized in natural polymers by gamma irradiation. Institute of Nuclear Chemistry and Technology, Warszawa 3. Cieśla K, Sartowska B, (2008) Scanning Electron microscopy studies of structural properties of films prepared using non-irradiated and gamma irradiated potato starch. Annual Report of the Institute of Nuclear Chemistry and Technology 2006, 49-52 4. Cieśla K, Sartowska B, Królak E, Głuszewski W(2007) Gamma irradiation influence on structure of potato starch gels studied by SEM. Annual Report of the Institute of Nuclear Chemistry and Technology 2006, 46-51.
204 Technological Developments
NANOPORES WITH CONTROLLED PROFILES IN TRACK-ETCHED MEMBRANES
B. Sartowska \ O. L. Orelovitch 2, A. Presz 3, P.Yu. Apel 2, I. V. Blonskaya 2
institute of Nuclear Chemistry and Technology, Warsaw, Poland 2 Flerov Laboratory of Nuclear Reactions, Joint Institute for Nuclear Research, Dubna, Russia 3Institute of High Pressure Physics of the Polish Academy of Science, Warsaw, Poland
Track membranes (TMs) are porous systems consisting of a polymer foil with thin channels - pores - from surface to surface. There is increasing interest in fabrication of nanopores in polymer films with pre-determined geometries. This interest is connected with development of nanoporous materials with unique properties such as diode-like pores in membranes, highly asymmetrical nanopores for molecular sensors and atom beam optics, development of nano- capillary bodies for modelling the transport of molecules and ions in constrained volumes. Control over pore geometry opens the way to a number of new applications of track-etch membranes. Surfactant - controlled etching method allows us to produce asymmetric track- etched nanopore membranes. The nanopores are fabricated by the ion-track etching method using of surfactant - doped alkaline solutions. By varying the alkali concentration in the etchant and the etching time, control over the pore profile and dimensions is achieved. In performing these studies, a key question is the detailed knowledge about geometrical characteristics of the nanoporous objects produced by ion-track technique. The pore geometry is characterised in detail using field-emission scanning electron microscopy. SEM images of the surfaces and cleavages of TMs with different pore morphology are shown.
205 NUTECH-2011
PRECISION LOW POWER X-RAY GENERATOR
T. Kotowski
Teleelektronika Zambrów, Poland
Precision low power X-ray generator was developed as a result of mutual work with Institute for Nuclear Studies Otwock-Swierk during experimental work over medical instrument using X-ray source in tumor brachytherapy, the so called Polish Photon Needle. In the initial stage this generator was developed to use it in industrial measurement instruments to replace source isotopes. It can be also used in other applications such as fluorescence analysis XRF, in studies and measurements where stable parameters of generated spectrum of radiation power to 60 keV are needed. The generator is adjusted to 100% duty and pulse cycle in range of tube voltage 3.500 V to 60.000 V and tube current 0,060 uA to 240,000 uA. The fundamental part of the generator is a miniature X-ray tube with beryllium window which is covered with transmission target. It can be constructed with tube containing target material Silver (Ag), Gold (Au), or Tungsten (W). In the tube strongly concentrated electron beam (50 um) is used and then adequate projection digital programme (PDPTT) on surface transmission target. Stability focusing in all range of changes of voltage current was solved (automatically) by compensation focusing system electron beam (CFSEB). One of the most important advantages of the equipment is high voltage stability + - 50 ppm/3 month, resolution high voltage - 1 V, resolution tube current - 10 nA , portion power - 1,8 mGy/h 60 kV, 70 uA (W target) from the source 30 cm, Efficiency low voltage (12 VDC input voltage) / high voltage - 82%. Basic limitations: low power source and absorption, low energy phantom of radiation X in broadcasting target. In the next stage, development and construction of high resolution, high efficiency and stability, cooled, Si-PIN crystal set is planned.
206 Technological Developments
PREPARATION OF 57Co SOURCES FOR MÖSSBAUER SPECTROSCOPY
I. Cieszykowska, M. Żółtowska, M. Mielcarski, A. Piasecki, T. Janiak, T. Barcikowski
Institute of Atomic Energy POLATOM, Radioisotope Center , 05-400 Otwock-Świerk
AIM. Cobalt-57 sources applied in Mössbauer Spectroscopy, utilizing the effect of recoilless gamma emission, are used in investigations of many processes proceeding in crystal lattice of solids. The method of preparation of 57Co source was developed in IAE POLATOM RC. METHOD. The preparation of Mössbauer sources comprised electrodeposition of carrier-free 57Co on rhodium foil [1] followed by thermal diffusion of 57Co into rhodium matrix [2], A series of experiments were performed in order to determine the optimal conditions for electrodeposition of cobalt on rhodium foils 6um thick. Electrochemical cell consisting of platinum anode and rhodium disc as the cathode was chosen. The electrolyte was an aqueous solution of ammonia citrate 25g/l, hydrazine hydrate 25/l and carrier- free 57Co in the form of 57Co(II) in 0,1M HCl. Ammonia solution (25%) was used for adjusting the pH to 10. The volumes of the electrolytes was 10ml and 5ml. The deposition rate 57Co was determined by measuring the gamma-activities of aliquot samples of the electrolyte before, during and after electrolysis in a well-type scintillation chamber. For comparison the same method was applied for determining the activity of the cathode after deposition under nearly similar geometrical conditions. The diffusion of 57Co into rhodium lattice was carried out by annealing the foil at temperature of 1100°C at high vacuum, in quartz tube. The 57Co active cores were encapsulated in cylindral Ti capsules with Be windows. The Mössbauer spectra were measured to verify the quality and efficiency of the testing sources. RESULTS. The experiments performed allow making an optimum choice of the electrodeposition parameters of carrier- free 57Co on rhodium foil. The highest efficiency approaching 100% and the best rate of deposition were obtained using current density 50mA/cm2 and electrolyte volume - 5ml. The best results of diffusion of electrodeposition cobalt-57 onto rhodium matrix was obtained in an annealing process at 1100°C in vacuum over 10"6 hPa. CONCLUSION. The main spectra parameters of the prepared sources are fairly acceptable with respect to the typical obtainable values for a-Fe absorbers in Mössbauer spectroscopy. The results obtained confirm that the deposited layer diffused almost completely into rhodium matrix without substantial loss of the activity deposited.
1. Cieszykowska I., Żółtowska M, Mielcarski M. (201 l)Electrodeposition of carrier-free 57Co on rhodium as an approach to the preparation of Mössbauer sources. Applied Radiation and Isotopes 69, 142-145 2. Cieszykowska I., Żółtowska M., Zachariasz P., Piasecki A., Janiak T., Mielcarski M. (2011) Thermal diffusion of 57Co into rhodium matrix as a second step in preparing Mössbauer Sources. Applied Radiation and Isotopes, doi: 10.1016/j.apradiso.2011.03.040
207 NUTECH-2011
PROMPT AND DELAYED GAMMA NEUTRON ACTIVATION ANALYSIS FOR THE ASSAY OF TOXIC ELEMENTS IN RADIOACTIVE WASTE PACKAGES
A. Havenith1, J. Kettler1, E. Mauerhofer2
e.mauerhofer@fz-juelich.
institute of Nuclear Fuel Cycle, RWTH-Aachen, D-52062 Aachen 2Institut for Energy and Climate Research, Nuclear Waste Management and Reactor Safety, Forschungszentrum Jülich GmbH, D-52428 Jülich
In addition to the radioactive components, radioactive waste packages may contain non radioactive hazardous and toxic substances such as heavy metals that can adversely affect humans and the environment [1]. The disposal of such radioactive waste packages must comply with appropriate regulations in order to avoid groundwater pollution. In Germany, the declaration and balancing of toxic substances in radioactive waste have become obligatory as result of the plan-approval for disposal "Konrad" [2], Thus, toxic elements in radioactive waste packages must be identified and quantified. An innovative analytical technique based on prompt and delayed gamma neutron activation analysis for non destructive assay of 200-l waste packages is described. Recent experimental and numerical results are presented.
1. Management of low and intermediate level radioactive wastes with regard to their chemical toxicity, IAEA-TECDOC-1325, December 2002. 2. Nieder sächsisches Umweltministerium; Planfeststellungsbeschuss für die Errichtung und den Betrieb des Bergwerkes Konrad in Salzgitter als Anlage zur Endlagerung fester oder verfestigter radioaktiver Abfälle mit vernachlässigbarer Wärmeentwicklung, t, Hannover, Nds. MBl. Nr. 21/2002.
208 Technological Developments
SCATTERED NEUTRON COMPONENT IN DIGITAL THERMAL NEUTRON RADIOGRAPHS OF SIMPLE OBJECTS
J. J. Milczarek1, F. C. de Beer2, M. J. Radebe2,1. M. Fijal-Kirejczyk1, J. Żołądek-Nowak1, A. Trzciński3
institute of Atomic Energy POL ATOM, Świerk, 05-400 Otwock, Poland, 2Radiation Science, Necsa, Church street west ext., Pelindaba, Pretoria, 0001, South Africa 3A. Soltan Institute for Nuclear Problems, Świerk, 05-400 Otwock, Poland
The obscuring effect of the scattered neutrons on the determination of the water content from digital thermal neutron radiographs of simple shape objects made of wet porous materials is of great importance in studies of liquid migration in porous media [1-3]. In this work the contribution of scattered neutrons to image formation of parallelepipeds, cylinders and wedges made of water was studied experimentally and by MC modelling. In order to compare the results from different experimental instruments two different neutron radiography stations were employed. The SANRAD facility at SAFARI research reactor and the NGRS facility at MARIA research reactor were used. In experiments the images of the exactly the same objects placed at presumed distances from the neutron to light converter screen were registered and analysed. The main aim of the work consisted in establishing practical rules for determination of the amount of water passed by neutron beam from the brightness of registered images and known object to detector distance. The distribution of brightness in the neutron radiographs were fitted with profiles determined with MC modelling. The results are compared to the analysis performed with QNI software package. In effect the effective macroscopic neutron cross section and the amplitude of scattered neutrons were determined. Some discrepancies between both approaches were found and explained in terms of different thermal neutron beam characteristics of both instruments employed. The nonuniformity encountered in distribution of water in porous specimens during water imbibition or drying was modelled by wedges made of water. The strong deviation of brightness profiles from linearity due to scattering neutrons was registered and quantitatively accounted for by results of MC simulations.
1. Deinert MR, Parlange JY, Steenhuis T, Throop J, Ünlü K, Cady KB (2004) Measurement of fluid contents and wetting front profiles by real-time neutron radiography. J. Hydrology 290: 192-201 2. Hassanein R, de Beer FC, Kardjilov N, Lehmann E (2006) Scattering correction algorithm for neutron radiography and tomography tested at facilities with different beam characteristics. Physica B 385-386: 1194-1196 3. Kaestner A, Hassanein R, Vontobel P, Lehmann P, Schaap J, Lehmann E, Flühler H (2007) Mapping the 3D water dynamics in heterogeneous sand using thermal neutrons. Chem Eng J. 130: 79-85
209 NUTECH-2011
STUDY OF ANGULAR DISTRIBUTION OF NEUTRON EMITTED FROM PLASMAS USING NUCLEAR REACTIONS INDUCED IN INDIUM S. Jednoróg1, A.Szydłowski2, M. Paduch1, M. Scholz1, B.Bieńkowska1, R.Prokopowicz1 [email protected] institute of Plasma Physics and Laser Microfusion Association EURATOM, Warsaw, 2Institute for Nuclear Studies, Association EURATOM, Swierk. Plasma produced in some contemporary fusion experiments is so hot and dense that nuclear reactions occur very intensively. For example at JET, the biggest tokamak all over the world, up to 1018 (d,t) reactions are induced per pulse in d-t discharges. Nuclear reactions taking place inside the plasmas are monitored effectively through their high energy product, mainly neutron detection. The neutrons are not only indicators of the nuclear reactions, but they carry away a lot of valuable information related to plasmas parameters. To extract this information sophisticated diagnostic methods are needed. At the IPPLM various neutron diagnostics methods have been development for many years and they are tested at FP-1000 facility [1] and implemented at other plasma experiments like JET[2] and, in the future at WX7. The neutron activation techniques have a special importance among others. This method supported by neutron transport calculations is the effective tool for thermonuclear plasma property investigation. Amid many activation materials indium occupied important position because it is relatively easily activated due to its high nuclear reaction cross section and it could selectively record neutrons with different energies. Among a few nuclear reactions induced by neutrons in an indium sample two of them are especially important. One is the radiation capture reaction In115(n,y)In116, which is threshold- less reaction unlike inelastic scattering reaction In115(n,n')In115m, which has a certain threshold energy. Separate studies have been carried out to optimize the measurement geometry and sample architecture (mass, and shape), which resulted in effective and efficient detections of neutrons [3]. The HPGe spectrometry has been used to measure the y-deexcitation of the indium samples in result of their activation at PF-1000 facility. Basing on this technique it was possible to estimate the contribution of neutrons directly coming from the plasma and discriminate them from scattered ones. The set of twenty indium samples has been used for measurement of angular distributions of neutrons around the vacuum chamber of PF-1000 facility. The indium samples were fixed on the outside wall of the discharge chamber and allow registration even small neutron yields because their geometry was optimized. The horizontal and azimuth distributions of neutrons were monitored from one discharge to the other. To measure the horizontal distribution the set of indium samples was distributed in the plane parallel to the electrode axis, meantime the azimuth measurements were performed with the indium samples located on the plane perpendicular to electrode axis respectively. The angular distributions of neutrons deliver very important information concerning nuclear reaction mechanisms. 1. Scholz M et al. (2007) Fast-neutron source based on Plasma-Focus device. In: Proceedings of the international Conference on Researches and Applications of Plasmas, Plasma 2007, Greisvald, Germany 2. Prokopowicz R et al.(2011) Measurements of neutrons at JET by means of the activation methods. Nuclear Instruments and Methods in Physics Research A 637: 119—127 3. Jednorog S et al.(2009) Numerical optimization of activation samples for the application of the activation technique to measure neutrons in large fusion devices like JET and ITER. 36th EPS Conference on Plasma Physics, Sofia, Bulgaria.(29th June 2009 - 3rd July 2009)
210 Technological Developments
STUDY OF NUCLEAR LEVEL DENSITIES FOR EXOTIC NUCLEI
M.N. Nasrabadi1 , M. Sepiyani
mnnasrabadi@ast. ui. ac.ir
Department of Nuclear Engineering, Faculty of Advanced Sciences & Technologies, University of Isfahan, Isfahan 81746-73441, Iran
Nuclear level density (NLD) is one of the properties of nuclei with widespread application including astrophysics and nuclear medicine. Since there has been little experimental and theoretical research on the study of nuclei which are far from stability line, studying NLD for these nuclei is of crucial importance. Also NLD is an important input for nuclear researches codes and hence studying methods for calculations this parameter is essential. In this work, various methods for calculating NLD for practical applications are presented.
1. Bethe HA, (1936) Phys. Rev, 50; 332 2. Gilbert A, Cameron AGW, (1965) Can. J. Phys. 43; 1446 3. Dilg W, Schantl W, Vonach H, Uhl M, (1973) Nucl. Phys. A 217; 269 4. IgnatyukAV, Istekov KK, Smirenkin GN, (1979) Sov. J. Nucl. Phys. 29: 4: 450 5. Goriely S, Tondeur F, Pearson JM, (2001) Atom. Data Nucl. Data Tables 77: 311 6. Goriely S, Hilaire S, Koning AJ, (2008) Phys. Rev. C 78; 064307 7. French JB, Ratcliff KF, (1971) Phys. Rev. C 3; 94 8. Hauser W, Feshbach H, (1952) Phys. Rev. 87; 366
211 NUTECH-2011
THE GROWTH AND SCINTILLATION CHARACTERISTICS OF LITHIUM DOPED CsI CRYSTALS
Maria da Conceição Costa Pereira, José Patrício Nahuel Cardenas and Tufic Madi Filho
Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP Av. Prof. Lineu Prestes 2242, 05508-000 - São Paulo - Brazil
Inorganic scintillators play an important role in the detection and spectroscopy of gamma and X-rays, as well as in neutrons and charged particles. For a variety of applications, new inorganic scintillation materials are being studied. New scintillation detector applications arise continuously and the interest in the introduction of new fast scintillators becomes relevant. Scintillation crystals based on cesium iodide (CsI) have relatively low hygroscope, easy handling and low cost, features that favor their use as radiation detectors. In this work, lithium doped CsI crystals were grown using the vertical Bridgman technique. In this technique, the charge is maintained at high temperature for 10 h to for the material melting and complete reaction. The temperature gradient 21° C/cm and 1 mm/h descending velocity are chosen as technique parameters. After growth is finished, the furnace is cooled at a rate of 20° C/h to room temperature. The concentration of the lithium doping element (Li) studied was 10-3 M. Analyses were carried out to evaluate the scintillator developed concerning two responses: a) to the gamma radiation, in the energy range of 350 keV to 1330 keV and b) to neutron from AmBe source, with energy range of 1MeV to 12 MeV. T.S. Korolevaa et al [1] describe in their paper about new scintillation materials, for registration of gamma-rays, X-rays, neutrons and neutrinos. One of these materials is 6Li. Lithium can capture neutrons without gamma-ray emission and, thus, reducing the back-ground. The neutron detection reaction is 6Li(n,a)3H with a thermal neutron cross section that 940 barns. In this paper we investigated the feasibility of the CsI:Li crystal as a gamma ray and neutron detector which can be used for monitoring, due to the fact that in our work environment we have two nuclear research reactors, calibration systems and radioisotope production.
1. Koroleva T S, Shulgin B V, Pedrini Ch, Ivanov V Yu, Raikov D V, Tcherepanov A N (2005). New scintillation materials and scintiblocs for neutron and y-rays registration. NuclInstrum Methods Phys Res 537: 415-423
212 Technological Developments
THE HPGE VIRTUAL POINT DETECTOR CONCEPT FOR RADIOACTIVE VOLUME-RING SOURCES BY MCNP4C SIMULATION
M.N. Nasrabadi
Department of Nuclear Engineering, Faculty of Advanced Sciences & Technologies, University of Isfahan, Isfahan 81746-73441, Iran
Validity of a virtual point detector model implying existence of a point where all interactions virtually occur was investigated for measurements of radioactive volume-ring sources. The correlations of the count rates with the distance between the virtual point detector and the detector face for radioactive volume-ring sources with various radii was studied by MCNP4c simulations and experimental data. Furthermore, the dependence of the virtual point detector on the y energy and sources with various radii was studied.
1. Presler O, German U, Pelled O, Alfassia ZB, (2004) The validity of the virtual point detector concept for absorbing media. Applied Radiation and Isotopes 60; 213-216 2. Alfassi ZB, Lavi N, Presler O, Pushkarski V (2007) HPGe Virtual point detector for radioactive disk sources. Applied Radiation and Isotopes 65; 253-258 3. Alfassia ZB, Pelled O, German U, (2006) The virtual point detector concept for HPGe planar and semi-planar detectors. Applied Radiation and Isotopes 64; 574-578. 4. Mahlinga S, Orion I, Alfassi ZB, (2006) The dependence of the virtual point-detector on the HPGe detector dimensions, Nuclear Instruments and Methods in Physics Research A 557; 544-553. 5. Presler O, German U, Pushkarskya V, Alfassia ZB, (2006) Virtual point detector on the interpolation and extrapolation of scintillation detectors counting efficiencies, Nuclear Instruments and Methods in Physics Research A 565; 704-710
213 NUTECH-2011
THE PERFORMANCE OF STRAW TUBES
S. Koperny, T. Z. Kowalski
AGH University of Science and Technology, Faculty of Physics and Applied Computer Science, Kraków, Poland
The straw tube is just a cylindrical drift tube having small diameter of cathode, typically from 4 to 8 mm. Straws are selected as the detecting element because they give the detector high modularity and compensate for failures. The granularity of the straw tube tracker provides up to 40 measurements point per track in current running detector, allowing effective track reconstruction. Due to their small dimension they can properly work at strong, varying magnetic fields and high radiation fields. There is no cross talk between anodes as was observed in a multiwire chamber because each straw is absolutely separate detecting element. Straws are mechanically robust. The following general straws properties will be described: • gas gain as a function of cylindrical high voltage for different counting gases, • longitudinal and transversal uniformity, • temperature and working gas pressure effects, • counting characteristics, • performance of straw in different outer environment.
214 Technological Developments
VERY LOW COST MULTICHANNEL ANALYSER WITH SOME ADDITIONAL FEATURES
K. Tudyka1 and A. Bluszcz1
1Centre of Excellence - Gliwice Absolute Dating Methods Centre, Institute of Physics, Silesian University of Technology, Gliwice, Poland
In this paper we present a multichannel analyser based on a digital signal controller (DSC). The multichannel analyser is characterized by a very low cost and an additional feature like recording time intervals between pulses. The total cost of electronic parts used in construction of the multichannel analyser is around USD 50. Electronic circuitry bases on dsPIC30F2020 digital signal controller unit from Microchip. The device has 10-bit Analogue to Digital Converters (ADC) which can sample and convert 2 samples per [is. The UART to USB communication is realized by MCP2200 from Microchip. The DSC is sampling the input voltage continuously and detects incoming pulses. Data belonging to a detected pulse and its time stamp are sent to a PC on-line. The analysis of data stored on the PC is performed off- line with the help of a genetic algorithm used to fit the pulse shape function. This allows to determine amplitude of each individual pulse. The effective resolution vary with the pulse length and is typically 1000 channels (10 bits) for pulses approximately 4 [is long. Preliminary results obtained on Canberra Quad Alpha Spectrometer 7404 are presented in Fig. 1. This project was directly inspired by [1].
250 n 60
200 ~ 50 ,40 150 -i I 30 100 J ' 20 50 -I 10 0 - 0 •Ai- 0 200 400 600 800 1000 ci 200 400 600 800 1000 channel channel 1000 n 120 100 800 :
600 : 400 -
200 0 J 0 2 4 6 8 10 50 100 150 time (ms) time (s) Figure 1. First results obtained with the described MCA. A & B - pulse amplitude spectra for the test signal and for alpha source respectively, C & D - distributions of time intervals for the test signal and for alpha source respectively.
1. Walter J, Barreiro M, Sajo-Bohus L, Greaves ED, Gonzalez W, 2005. New approach in add-on multi-channel analyser for gamma ray spectrometry. Nuclear Instruments and Methods A 545:776-783.
215 NUTECH-2011
WIRELESS SYSTEM FOR RADIOMETRIC MEASUREMENTS
A. Jakowiuk1, B. Machaj, P. Pieńkos, E. Kowalska, P. Filipiak, E. Świstowski
institute of Nuclear Chemistry and Technology, 16 Dorodna Streat, 03-195 Warsaw
Wireless system for radiometric measurements contains probes for gamma radiation measurements, and other probes for radon concentration measurements in air and in water. The probes have the form of droplet-tight cylinders powered from a local battery. Measuring data collecting unit, based on portable computer, communicates directly with the probes in wireless manner using the Wi-Fi communication network, or through Internet using mobile phone GSM network. Serial port wire connection is also possible. The local battery ensures at least 14 days continuous operation of the probes. For long term measurements and when more measuring probes are used in investigations carried out, the probes can also be powered from solar panels. The wireless system for radiometric measurements includes: • Measuring probes for radon concentration measurement in air using natural diffusion of radon into Lucas cell through a light tight membrane placed at the perforated top of the cell opposite the cell window.
• Measuring probes for radon concentration in water using standard Lucas cell with two stubs connected with elastic pipes to a membrane in form of a pipe up to 4 m long, external diameter 5 mm, wound around perforated cylinder 11 cm in diameter making a spiral form of the membrane. The membrane is immersed in water. Radon from water diffuses into the membrane. A small air pump of the probe forces radon in the membrane radon to flow through the Lucas cell in closed loop. Water temperature sensor is also immersed in water.
• Measuring probes for measurement gamma radiation using NaI(Tl) scintillator, c[)50x50 mm coupled to photomultiplier tube c])50 mm (scintillation integrated detector)
216 Technological Developments
MODELLING OF FUEL EQUILIBRIUM IN LEAD-COOLED FAST REACTORS
J. Cetnar, G. Domańska, P. Stanisz
AGH University of Science and Technology, Krakow
Lead-cooled Fast Reactors (LFR) belong to Generation IV of nuclear reactors and significantly differ from ones of previous generation in regards to fuel cycle and achievable burnup. Presented works has been performed within LEADER project of EURATOM 7th FP. As any fast reactor it can be design to be a net fuel breeder (the fuel breeding to exceed unity), but in the current project other goals are aimed at. They are fuel self-breeding (just for sustaining its own fuel cycle) with optional consumption of additional amount of minor actinides (MA). The fuel self-breeding option bring some consequences on the core performance over time, which depend on designed fuel zoning and the system of reactivity control and compensation. Many factors come into play, as neutron flux has its buckling that adversely changes power distribution in possible power peaking formation, and occurrence of reactivity swing might require control rod insertion adjustment thus influencing power distribution further. In self-breeding option the fuel undergoes both burnup and transmutation and after its recycling with addition of fresh fertile material, converts toward its equilibrium composition. For the sake of fuel cycle efficiency it is desirable that the burnup is sufficiently high - within the limits of radiation robustness, and uniform on discharge. To achieve that goal, different options of the core arrangement has been analyzed that include location of the control rods - shutdown and compensation ones, different schemes of fuel reloading as well as few options of fuel enrichment zoning. Description of advanced modelling methodology for fuel cycle equilibrium will be given with quantitative comparison to the simplified approach. Calculations were performed using Monte Carlo system MCB5 based on MCNP5 recently equipped with error propagation modules and step predictor-corrector features.
217