Distribution of residual long-lived radioactivity in the inner concrete walls of a compact medical cyclotron vault room (original) (raw)
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Radioactivation Analysis of Concrete Wall in OKTAVIAN Facility
Plasma and Fusion Research
A deuterium-tritium (DT) neutron generator in Osaka University with a continuous intense neutron source emitting 3 × 10 12 fusion neutrons per second has been in operation since 1981. However, radioactivation for the parts of the accelerator body is a serious issue. Hence, in this study, we investigated the radioactivation of the intense irradiation room containing the continuous intense neutron source. Core samples of the concrete wall were collected at various positions in the irradiation room and the radionuclides in them were determined by performing gamma-ray spectrometry. Major long-lived radionuclides found were 54 Mn, 60 Co, and 152 Eu. The radioactivity of 152 Eu may possibly be consistent with the result obtained using the simulation code. The radioactivities of 54 Mn and 60 Co were minimal compared with that of 152 Eu. The tritium amount in the core sample was measured employing a tritium sampling system and a liquid scintillation detector and was found to be considerably larger than the amount estimated using the simulation code. Tritium diffused from the titanium-tritium target was attached to the wall surface. However, most of it did not penetrate the concrete wall. These results reveal the radioactivity issue of fusion neutron generator facilities and are expected to aid in the maintenance of their operation.
Medical Physics, 2010
Purpose:The authors want to assess the relevance of the neutron activation of the concrete vault of the PET cyclotron at CIMES (Universidad de Malaga) by predicting specific activities of the main activation products in the vault and their variation profiles as a function of penetration depth into concrete at present and after 10 yr of cyclotron operation.Methods:The dual proton cyclotron is used for PET isotopes production, mainly. During the years 2006 and 2008, the using rate has been 1 h/day at single beam . From January 2008, using rate is 4 h/day at dual beam . The energy of the cyclotron proton beam is 18 MeV. Four point locations were chosen on the walls of the cyclotron room to assess neutron induced activity concentrations. In each wall point location, neutron induced radionuclide specific activity was assessed from the wall surface to a depth of 120 cm within concrete. Simulations were carried out with the Monte Carlo based radiation transport code MCNPX (v2.6.0).Results:...
Study on Concrete Activation Reduction in a PET Cyclotron Vault
Journal of Radiation Protection and Research, 2020
Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd2O3, B4C, polyethylene (PE), and borated (5 wt% nat B) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B4C and Gd2O3 were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1-or 0.4-cm-thick B4C or Gd2O3 behind 10 cmthick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cmthick Gd2O3+10 cm-thick PE or 0.1 cm-thick B4C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B4C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.
Radiochemical analysis of concrete samples from accelerator waste
Radiochimica Acta, 2012
For the decommissioning and disposal of shielding concrete from accelerator facilities, the Swiss Authorities require information on the radionuclide inventory. Besides the easy-to-measure γ-emitters 152Eu, 60Co, 44Sc, 133Ba, 154Eu, 134Cs, 144Ce, 22Na, also long-lived radionuclides emitting α- or β-radiation like 129I, 10Be, 36Cl, 239/240Pu and 238U have to be studied in order to obtain an overview to which extent they are produced and whether they represent a safety issue. In this study, we present the chemical separation and determination of selected radionuclides in shielding concrete from two different positions in the accelerator facilities at the Paul Scherrer Institute (PSI), the BX2 station, which was shut down in 1998, and the environment of the target M station, where the samples were taken in 1985 during reconstruction. The results of the measurements show that in no case the radionuclide content represents a safety risk. The components can be decommissioned corresponding...
The Radioactivity Estimation of the Irradiated 13 Mev Cyclotron’s Concrete Shield
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
The Center for Accelerator Science and Technology (PSTA) planned to install K500 concrete shield in its 13 MeV cyclotron facility (DECY-13). However, fast neutrons that are generated by this cyclotron could activate materials of the concrete. It may harm the radiation workers. In this work, we conducted simulations using ORIGEN2 and PHITS computer code to estimate the formed radioactivity and the neutron flux distribution in the DECY-13 cyclotron's concrete shield. Based on the simulation, the induced radioactivity is 2.3478 × 109 Bq, while its gamma dose rate is 22.09 µSv/m2h. The most contributed isotopes are Th-233, Ho-166, Al-28, Mn-56 and Si-31. This dose is quite high. Neutron fluxes in the rear of the simulated concrete shield are also still prominent. Accordingly, it is necessary to attach neutron shielding materials which do not generate high-intensity gamma-ray. The formed radioactivity is high; but it appears from the short half-life isotopes such as Th-233, Ho-166, A...
Journal of Radioanalytical and Nuclear Chemistry, 2020
Cement is an important component of concrete used as a shielding material in nuclear accelerators and reactors. Hence cement samples should be analysed for the presence of certain trace elements that may get activated by neutrons emitted during the production of radioisotopes in an accelerator, so as to minimize the low level radioactive waste to be handled during decommissioning. With this motivation the present work was undertaken and 44 samples of five broad classes of cements were analysed for natural radioactivity (226 Ra, 232 Th & 40 K) and trace elements capable of generating long lived gamma radioactivity due to neutron activation.
DETERMINATION OF THE RADIOACTIVITY LEVEL OF CONCRETE USED AS SHIELDING FOR MEDICAL 60Co SOURCE
RAP Conference Proceedings
This study examines the natural and artificial radioactivity in concrete used as shielding material for medical 60 Co source temporary stored in our waste storage site. The determination of the radioactivity level is done to see if any leakage or contamination occurred in concrete material after the dislocation of 60 Co source to another destination. Concrete samples were taken from the three drums located in the temporary waste storage site and after preparation of samples were placed in a marinelli beaker with a volume of 500 ml and left in isolation for one month to achieve the secular equilibrium. The activity concentrations of 40 K, 226 Ra and 232 Th in ten samples are determined by using gamma-ray spectrometry method with HPGe detector. The average values of activity concentration are found to be 147.56 ± 6.97 Bqkg-1 for 40 K, 18.09 ± 0.64Bqkg-1 for 226 Ra and 16.90 ± 0.68 Bqkg-1 for 232 Th, respectively. The activity concentration index (ACI) is used as a screening tool to assess the radiological hazard due to possible release of the concrete in environment or to reuse it as building materials. From all analysis performed the maximum value of ACI was 0.21. This value was found to be lower than 1 and in none of them was found the presence of 60 Co radionuclide. We conclude depending on the Decision No. 638, dated on 07.09.2016 on the approval of the regulation "On the safe management of radioactive waste in the Republic of Albania" that the concrete could be discharged freely in environment, or it can be used as building material because do not pose any significant risk to humans.
Residual radioactivity at the CERN 600MeV synchro-cyclotron
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2012
The 600 MeV synchro-cyclotron (SC) was the first accelerator that came into operation at CERN in 1957. It provided beams for CERN's first particle and nuclear physics experiments and operated for 33 years until it was shut down in 1990. In view of a planned partial decommissioning of the facility, a range of measurements were carried out to evaluate the levels of residual radioactivity in the accelerator and its surrounding after about 20 years of cooling time. Gamma spectrometry measurements were performed on 113 samples collected inside the three floors of the accelerator vault, on the cyclotron itself and on concrete samples taken from various parts of the building walls, up to a depth of 50 cm in the shield. About 40% of all samples contain traces of neutron-induced radionuclides, mainly 60 Co (in metals), 133 Ba, 137 Cs, 152 Eu and 154 Eu (in concrete). Values of specific activities range from 5 mBq/g to 781 Bq/g. The maximum activity induced in concrete was observed at the depth of 40 cm in the wall near the cyclotron extraction channel. The laboratory measurements were supplemented by in-situ gamma spectrometry performed with the ISOCS system. A complete dose rate survey was also performed yielding isodose maps of the three levels of the building. The isotope production and the residual radioactivity in the barite walls of the SC bunker were simulated with the FLUKA and JEREMY codes in use at CERN for predicting residual radioactivity in activated accelerator components, and the results compared with the gamma spectrometry data. A detailed comparison of calculated and measured specific activities shows generally good agreement, to within a factor 2 in most cases. These results serve as indirect validation of the capabilities of these codes to correctly predict residual radioactivity with only a very approximate knowledge of the irradiation profile and after a very long (20 years) cooling time. Overall the results provided in this paper may be of use for estimating residual radioactivity in proton accelerators of comparable energy and for benchmarking computer codes.
Assessment of long-lived residual radioisotopes in cement induced by neutron radiation
MATEC web of conferences, 2020
During the decommissioning of nuclear power plants, a significant amount of cement based composites should be disposed as radioactive waste. The use of material with low-activation constituents could effectively reduce radioactivity of concrete. The subject of the paper is the content of trace elements with large activation cross section in concrete constituents due to their ability to be activated in radiation shielding structures. Various Portland cement specimens were subjected to elemental analysis by neutron activation analysis and prompt gamma activation analysis to assess the dominant long-lived residual radioisotopes. Concentrations of the radionuclides, such as Europium-152, Cobalt-60 and Caesium-134 were assessed. Their half-life time is 13.5, 5.27, and 2.07 years, respectively. On the basis of the obtained results, recommendations for cement selection for low-activation concrete are proposed in order to economize decommissioning cost by reducing a radioactive concrete waste.