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1
Content available Praca reaktora badawczego MARIA w 2022 roku
Witkowski Piotr
Postępy Techniki Jądrowej
|
2023
|
z. 1
2--8
PL
Wysokostrumieniowy reaktor badawczy MARIA, eksploatowany w Narodowym Centrum Badań Jądrowych w Świerku, wykorzystywany jest do produkcji radioizotopów oraz do prowadzenia badań z wykorzystaniem wiązek neutronów. W artykule opisano parametry techniczne reaktora i charakterystykę jego pracy w 2022 r.
EN
The MARIA high-flux research reactor operated at the National Centre for Nuclear Research at Swierk (Poland) is used for targets irradiation and to run physical experiments. The technical parameters of the reactor and characteristics of its operation in 2022 are presented.
2
Content available remote Reactor Laboratory for Biomedical Research in National Centre for Nuclear Research in Poland
Michaś Edyta, Dorosz Michał, Wójciuk Karolina, Tymińska Katarzyna, Wiliński Maciej, Maciak Maciej, Domański Szymon, Wojtania Grzegorz, Bartosik Łukasz, Małkiewicz Adam, Lechniak Jan, Gryziński Michał Aleksander
Polish Journal of Medical Physics and Engineering
|
2021
|
Vol. 27, Iss. 1
119--122
EN
National Centre for Nuclear Research, NCBJ is one of the biggest research institutes in Poland, in which scientists deal with basic research in the various fields of subatomic physics, development of nuclear technologies and practical applications of nuclear physics methods, including those for nuclear medicine and radiotherapy. NCBJ operates the only Polish nuclear research reactor MARIA, around which a Reactor Laboratory for Biomedical Research, RLBR has been built in the last 4 years. One of the main aims of the RLBR team is to adapt the H2 channel, one of the eight MARIA’s horizontal channels, to a specific irradiation facility delivering a high flux thermal/epithermal neutron beam. The beam derived from the channel will be a tool for biological, physical and material studies for Boron Neutron Capture Therapy, BNCT. While NCBJ is focused on building a neutron research facility, the Polish scientific community expressed its interest in BNCT development and implementation as an alternative therapy for cancer treatment. Through the working group meetings organized in the form of regular scientific workshops since 2015, it led to the establishment of a national scientific consortium dedicated to BNCT. Polish Consortium for Boron Neutron Capture Therapy agreement was initially signed by twelve institutions including scientific institutes, universities and oncological centres in October 2019. National Centre for Nuclear Research was appointed the leader of the consortium. A year later the consortium was enlarged by two more institutions.
3
Content available Source term estimation for the MARIA research reactor and model of atmospheric dispersion of radionuclides with dry deposition
Lipka Maciej
Nukleonika
|
2020
|
Vol. 65, No. 3
173--179
EN
Source term is the amount of radionuclide activity, measured in becquerels, released to the atmosphere from a nuclear reactor, together with the plume composition, over a specific period. It is the basis of radioprotection- -related calculation. Usually, such computations are done using commercial codes; however, they are challenging to be used in the case of the MARIA reactor due to its unique construction. Consequently, there is a need to develop a method that will be able to deliver useful results despite the complicated geometry of the reactor site. Such an approach, based upon the Bateman balance equation, is presented in the article, together with the results of source term calculation for the MARIA reactor. Additionally, atmospheric dispersion of the radionuclides, analysed with the Gauss plume model with dry deposition, is presented.
4
Content available Prace modernizacyjne w reaktorze Maria
Mikulski A.
Postępy Techniki Jądrowej
|
2018
|
z. 3
33--37
PL
Przedstawiono przeprowadzone prace modernizacyjne w reaktorze MARIA w czasie jego prawie 45-letniej eksploatacji. Można stwierdzić, że większość urządzeń istotnych dla jego bezpiecznej eksploatacji została wymieniona. Z pierwotnego wyposażenia pozostał tylko zbiornik, konstrukcja podtrzymująca elementy paliwowe i rurociągi obu obiegów.
EN
The majority of refurbishment work in the MARIA reactor done during its almost 45 years of operation is presented. The most of devices important for its safe operation was exchanged. From the initial equipment only reactor pool, construction supporting fuel elements and pipes of both cooling circuits are the same.
5
Content available Investigation of 99Mo potential production via UO2SO4 liquid target irradiation in a 5 MW nuclear research reactor
Gholamzadeh Z., Mirvakili S. M., Davari A., Alizadeh M., Joz-Vaziri A.
Nukleonika
|
2017
|
Vol. 62, No. 1
37--45
EN
The activation method for 99Mo production in comparison to fi ssionable target irradiation in research reactors is less preferable. Therefore, 99Mo yield using UO2SO4 samples was theoretically investigated. Computational results revealed admirable potential of the liquid samples for 99Mo production. Low-concentrated uranyl sulphate samples could easily be handled by the irradiation box. The sample geometry optimization improves thermal hydraulic conditions and production yield. The optimized geometry including only 0.12 g 235U produced 57Ci99Mo at end-of-irradiation (EOI) with a temperature peak of 72°C during the irradiation.
6
Content available Production of fission product 99Mo using high-enriched uranium plates in Polish nuclear research reactor MARIA: Technology and neutronic analysis
Jaroszewicz J., Marcinkowska Z., Pytel K.
Nukleonika
|
2014
|
Vol. 59, No. 2
43--52
EN
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.
7
Content available RELAP5/MOD3 model and transient analyses for the MARIA research reactor in Poland
Szczurek J., Czerski P., Bykowski W.
Nukleonika
|
2004
|
Vol. 49, nr 4
149-157
EN
The RELAP5/MOD3 input data model of the MARIA research reactor has been developed to provide the capability for the analysis of the reactor core under loss of flow and reactivity insertion transients. The model was qualified against the reactor data at steady state conditions and, additionally, against the existing reliable experimental data for a transient initiated by the reactor scram. The results obtained with the code agree well with the experimental data. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. The presented input data model should be treated as the first step for developing of the model including the whole primary cooling circuit of the reactor.
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