Janusz Jaroszewicz, Zuzanna Marcinkowska and Krzysztof Pytel
The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online 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 neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium 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.
Stanisław Kilim, Elżbieta Strugalska-Gola, Marcin Szuta, Marcin Bielewicz, Sergej I. Tyutyunnikov, Walter I. Furman, Jindra Adam and Vladimir I. Stegailov
natural uranium target assembly QUINTA irradiated by deuterons with energies from 1 to 8 GeV at JINR NUCLOTRON. XXI International Baldin Seminar on High Energy Physics Problems, 10-15 September 2012, JINR, Dubna, Russia. PoS(Baldin ISHEPP XXI)086.
5. Evaluated Nuclear Data File (ENDF). Interpreted ENDF file. NP-237(FY_cum)NP-238,RNP MAT=9346 MF=8 MT=459 Library: ENDF/B-VII.1. (n,FY_cum) Cumulative Fission-Product Yields and (n,ind_FY) Independent Fission-Product Yields. Retrieved from https://www-nds.iaea.org/exfor.endf.htm.
6. Table of
Stanislav Yu. Melchakov, Dmitry S. Maltsev, Vladimir A. Volkovich, Leonid F. Yamshchikov, Dmitry G. Lisienko, Aleksandr G. Osipenko and Mikhail A. Rusakov
1. Kinoshita, K., Kurata, M., & Inoue, T. (2000). Estimation of material balance in pyrometallurgical partitioning process of transuranic elements from high-level liquid waste. J. Nucl. Sci. Technol., 37(1), 75-83.
2. Moriyama, H., Kinoshita, K., Asaoka, Y., Moritani, K., & Ito, Y. (1990). Equilibrium distributions of actinides and fissionproducts in pyrochemical separation systems (II) LiCl-KCl/Cd system. J. Nucl. Sci. Technol., 27(10), 937-943. DOI: 10.1080/18811248.1990.9731273.
3. Lebedev, V. A
Przemysław Stanisz, Jerzy Cetnar and Grażyna Domańska
Report LEADER-DEL 005-2011, WP2, LEADER Project).
5. X-5 Monte Carlo Team. (2003). MCNP - A General Monte Carlo N-Particle Transport Code, Version 5. Los Alamos: Los Alamos National Laboratory. (LAUR-03-1987).
6. Cetnar, J. (2006). General solution of Bateman equations for nuclear transmutations. Ann. Nucl. Energy, 33(7), 640-645.
7. Cetnar, J., Gudowski, W., & Wallenius, J. (1999). MCB: A Continuous Energy Monte Carlo Burnup Simulation Code. Actinide and FissionProduct Partitioning and Transmutation. (EUR 18898 EN, OECD
Tomasz Pliszczyński, Katarzyna Ciszewska, Małgorzata Dymecka, Jakub Ośko and Zbigniew Haratym
Fission products of 235U or isotopes from activation may appear in technological waters at normal operation of a research reactor. Therefore, reactor cooling water may contain a number of beta radioactive isotopes including yttrium and strontium isotopes, which can pose potential hazard to reactor personnel. In order to asses internal exposure urinalysis is carried out. This work presents the method of sample preparation and measurement used by Radiation Protection Measurements Laboratory (RPLM) of the National Centre for Nuclear Research (NCNR). Method of various isotopes of yttrium and Sr-90 activity calculation is also shown. Determination of these isotopes activities in urine sample allows estimating the effective doses
Davide Rodrigues, Gabriela Durán-Klie and Sylvie Delpech
The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR) concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.
Nikolay Uzunov, Galina Yordanova, Seniha Salim, Natalya Stancheva, Vanya Mineva, Laura Meléndez-Alafort and Antonio Rosato
th edn., Maisonneuve, Sainte-Ruffine
. Council of Europe (2005b) Sodium pertechnetate [99mTc] injection (nonfission). Monograph No. 283. European Pharmacopoeia, 5th edn. Maisonneuve, Sainte-Ruffine
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1. Cetnar, J., Gudowski, W., & Wallenius, J. (1999). MCB: A continuous energy Monte Carlo burn-up simulation code. In Proceedings of Actinide and FissionProduct Partitioning and Transmutation. (EUR 18898 EN, OECD/NEA 523).
2. Suyama, K., Murazaki, M., Ohkubo, K., Nakahara, Y., & Uchiyama, G. (2011). Re-evaluation of assay data of spent nuclear fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems. Ann. Nucl. Energy, 38, 930-941. DOI: 10.1016/j.anucene.2011
Natalia Szewczuk-Krypa, Marta Drosińska-Komor, Jerzy Głuch and Łukasz Breńkacz
1. Baek J-K., Mistarihi Q., Yeo S., et al.: A Preliminary Study for Diffusion Experiments of Metallic FissionProducts in Graphite for HTGR . Transaction of the Korean Nuclear Society Autumn Meeting Gyeongiu, Korea, October 29-30, 2015
2. Carlton J S., Smarta R., Jenkins V.: The nuclear propulsion of merchant ships: Aspects of engineering, science and technology . Journal of Marine Engineering & Technology 2011, pp. 47-59
3. Gardzilewicz A., Głuch J., Bogulicz M. (1994): DIAGAR manual for turbine set No. 3 at Kozienice Power Plant
Slawomir Jednorog, Ewa Laszynska, Barbara Bienkowska, Adam Ziolkowski, Marian Paduch, Kamil Szewczak, Katarzyna Mikszuta, Karol Malinowski, Marcel Bajdel and Pawel Potrykus
9. Bhatia, C., Fallin, B., Gooden, M. E., Howell, C. R., Kelley, J. H., Tornow, W., Arnold, C. W., Bond, E. M., Bredeweg, T. A., Fowler, M. M., Moody, W. A., Rundberg, R. S., Rusev, G., Vieira, D. J., Wilhelmy, J. B., Becker, J. A., Macri, R., Ryan, C., Sheets, S. A., Stoyer, M. A., & Tonchev, A. P. (2014). Dual-fission chamber and neutron beam characterization for fissionproduct yield measurements using monoenergetic neutrons. Nucl. Instrum. Methods Phys. Res. Sect. A-Accel. Spectrom. Detect. Assoc. Equip ., 757 , 7–19. DOI: 10.1016/j.nima.2014