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Investigation of 99Mo potential production via UO2SO4 liquid target irradiation in a 5 MW nuclear research reactor

Monte Carlo serpent analysis of the HOR research reactor and its medical isotope production capabilities using uranium salts. Thesis, Delft University of Technology, The Netherlands. 9. Micklich, B. J. (2015). Remanent activation in the mini-SHINE experiments. In 3rd International Workshop on Accelerator Radiation Induced Activation (ARIA’15), 15–17 April 2015, Knoxville, Tennessee, USA (36 pp.). Available from https://public.ornl.gov/neutrons/conf/aria2015/presentations/12%20Remanent%20Activation%20in%20the%20mini-SHINE%20Experiments.pdf . 10. May, I

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Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

Abstract

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.

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Modelling of thermal hydraulics in a KAROLINA calorimeter for its calibration methodology validation

Abstract

Results of numerical calculations of heat exchange in a nuclear heating detector for nuclear reactors are presented in this paper. The gamma radiation is generated in nuclear reactor during fission and radiative capture reactions as well as radioactive decay of its products. A single-cell calorimeter has been designed for application in the MARIA research reactor in the National Centre for Nuclear Research (NCBJ) in Świerk near Warsaw, Poland, and can also be used in the Jules Horowitz Reactor (JHR), which is under construction in the research centre in Cadarache, France. It consists of a cylindrical sample, which is surrounded by a gas layer, contained in a cylindrical housing. Additional calculations had to be performed before its insertion into the reactor. Within this analysis, modern computational fluid dynamics (CFD) methods have been used for assessing important parameters, for example, mean surface temperature, mean volume temperature, and maximum sample (calorimeter core) temperature. Results of an experiment performed at a dedicated out-of-pile calibration bench and results of numerical modelling validation are also included in this paper.

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Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

N-Particle Transport Code System . (RSICC code package CCC-700/MCNP4C). Oak Ridge National Laboratory, TN, and DOE, USA. 5. Jonah, S. A., Liaw, J. R., & Matos, J. E. (2007). Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1). Ann. Nucl. Energy , 34 , 953–957. 6. Khattab, K., & Sulieman, I. (2009). Calculations of the thermal and fast neutron fluxes in the Syrian MNSR irradiation tubes using the MCNP-4C code. Appl. Radiat. Isot ., 67 , 535–538. 7. Balogun, G. I. (2003). Automating some analysis and

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Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

. Des., 180, 133-146. 4. Usha, S., Ramanarayanan, R. R., Mohanakrishnan, P., & Kapoor, R. P. (2006). Research reactor KAMINI. Nucl. Eng. Des., 236, 872-880. 5. Kumar, A., Srivenkatesan, R., & Sinha, R. K. (2009). On the physics design of advanced heavy water reactor (AHWR). In International Conference on Opportunities and Challengers for Water Cooled Reactors in the 21st Century, 27-28 October 2009 (pp. 84-85). Vienna: International Atomic Energy Agency. (IAEA-CN-164). 6. Maitra, R. (2005) Thorium: Preferred nuclear fuel

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Concept of a BNCT line with in-pool fission converter at MARIA reactor in Swierk

References Golnik N, Pytel K, Dabkowski L. A concept and state of the art. of irradiation facilities for NCT at research reactor MARIA in Poland. In: Sauerwein W, Moss R, Witting A, editors. Research and Development in Neutron Capture Therapy. Monduzzi Editore; 2002. p. 191-195. Golnik N, Pytel K. Irradiation facilities for BNCT at research reactor MARIA in Poland. Pol J Med Phys Eng. 2006; 12(3): 143-153. Krzysztoszek G, Golab A, Jaroszewicz J. Operation of the MARIA research

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A Novel Control-rod Drive Mechanism via Electromagnetic Levitation in MNSR

. , 85 , 655-661. 7. International Atomic Energy Agency. (1986). Technology and use of low power research reactors . Vienna: IAEA. (IAEA-TECDOC-384). 8. Tabadar, Z., Hadad, K., Nematollahi, M. R., Jabbari, M., Khaleghi, M., & Hashemi-Tilehnoee, M. (2012). Simulation of a control rod ejection accident in a VVER-1000/V446 using RELAP5/Mod3.2. Ann. Nucl. Energy , 45 , 106-114. 9. Ku, C. L., Li, T. H. S., & Guo, N. R. (2005). Design of a novel fuzzy sliding-mode control for magnetic ball levitation system. J. Intell. Robot. Syst. , 42 (3), 295-316. 10

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Estimation of Control Rod Worth in a VVER-1000 Reactor using DRAGON4 and DONJON4

References 1. Lee, E. K., Shin, H. C., Bae, S. M., & Lee, Y. K. (2005). New dynamic method to measure rod worths in zero power physics test at PWR startup. Ann. Nucl. Energy , 32 , 1457-1475. 2. Shimazu, Y., Okazaki, K., & Tsuji, M. (2006). Feasibility study for evaluation of control rod worth in pressurized water reactors using neutron count rate during a control rod drop testing. Nucl. Sci. Technol. , 43 , 919-923. 3. Kalcheva, S., & Koonen, E. (2009). Improved Monte Carlo-Perturbation method for estimation of control rod worths in a research

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Studies on hydrometallurgical processes using nuclear techniques to be applied in copper industry. II. Application of radiotracers in copper leaching from flotation tailings

radiotracer for investigation of copper ore leaching. Nukleonika, 63(4), 123-129. DOI: 10.2478/nuka-2018-0015. 24. Bujdoso, E., Feher, I., & Kardos, G. (1973). Activation and decay tables of radioisotopes. Amsterdam, New York: Elsevier. 25. Jaroszewicz, J., Marcinkowska, Z., & Pytel, K. (2014). Production of fi ssion product 99Mo using high-enriched uranium plates in Polish nuclear research reactor MARIA: Technology and neutronic analysis. Nukleonika, 59(2), 43-52. DOI: 10.2478/nuka-2014-0009. 26. Chmielewski, T. (2016

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Assessment of occupational internal exposure to beta emitters from the nuclear reactor primary coolant circuit

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

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