Determination of the radionuclide inventory in accelerator waste using calculation and radiochemical analysis (original) (raw)
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Radiochemical analysis of a copper beam dump irradiated with high-energetic protons
Radiochimica Acta, 2009
The radionuclide inventory of a copper beam dump from the 590 MeV proton accelerator of the Paul Scherrer Institute in Switzerland was determined, focusing on radioisotopes with half-lives of more than 60 d, and in particular, of long-lived isotopes with T 1/2 = 10 4-10 7 years, which are important regarding radioactive waste management. The measurements were carried out using high resolution γ-measurement without sample destruction, as well as liquid scintillation counting (LSC) and accelerator mass spectrometry (AMS) after chemical separation. For the first time, a beam dump from a high power accelerator facility was completely characterized concerning the depth and radial distribution profile of the most hazardous and/or long-lived radionuclides. Moreover, it turned out that some of the investigated radionuclides, like for instance 26 Al, 44 Ti or 60 Fe represent valuable material for application in several scientific fields like nuclear astrophysics, basic nuclear physics research, radiopharmacy and many others. Therefore, based on the analytical results, a special research and development program has been started at PSI objecting on specific preparative extraction of longlived radioisotopes (ERAWAST-exotic radionuclides from accelerator waste for science and technology).
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...
EPJ Web of Conferences, 2017
The Paul Scherrer Institute (PSI) is the largest national research center in Switzerland. Its multidisciplinary research is dedicated to a wide field in natural science and technology as well as particle physics. The High Intensity Proton Accelerator Facility (HIPA) has been in operation at PSI since 1974. It includes an 870 keV Cockroft-Walton pre-accelerator, a 72 MeV injector cyclotron as well as a 590 MeV ring cyclotron. The experimental facilities, the meson production graphite targets, Target E and Target M, and the spallation target stations (SINQ and UCN) are used for material research and particle physics. In order to fulfill the request of the regulatory authorities and to be reported to the regulators, the expected radioactive waste and nuclide inventory after an anticipated final shutdown in the far future has to be estimated. In this contribution, calculations for the 20 m long beamline between Target E and the 590 MeV beam dump of HIPA are presented. The first step in the calculations was determining spectra and spatial particle distributions around the beamlines using the Monte-Carlo particle transport code MCNPX2.7.0 [1]. To perform the analysis of the MCNPX output and to determine the radionuclide inventory as well as the specific activity of the nuclides, an activation script [2] using the FISPACT10 code with the cross sections from the European Activation File (EAF2010) [3] was applied. The specific activity values were compared to the currently existing Swiss exemption limits (LE) [4] as well as to the Swiss liberation limits (LL) [5], becoming effective in the near future. The obtained results were used to estimate the total volume of the radioactive waste produced at HIPA and have to be reported to the Swiss regulatory authorities. The comparison of the performed calculations to measurements is discussed as well.
Health Physics, 2007
The LINAC-ADONE accelerator complex of the INFN-LNF Frascati National Laboratories, operating for 27 y prior to the commissioning of DA⌽NE, was dismantled in 1993. The scraps resulting from the decommissioning of LINAC-ADONE have been temporarily stored in the same Frascati laboratory, waiting for definitive disposal. Relying on recommendations of the IAEA, European Commission and Italian committees, an experimental characterization study of the LNF repository was performed. The main objective was a classification of the scraps on the basis of internationally recognized "clearance levels," which are 0.1 Bq g ؊1 for the isotopes of interest for this work. Secondly, a measurement of the materials suspected to be above 0.1 Bq g ؊1 was planned. Activation isotopes were expected from the aluminum, copper, steel, and iron of the LINAC and the ADONE ring sections. For screening purposes, the repository area has been divided into zones, where in-situ measurements with a portable HP-Ge detector have been performed. In addition, small samples have been cut from a representative number of pieces, and accurate laboratory measurements have been made with a low background HP-Ge spectrometer. The experimental results are in good agreement with other studies and show that a large part of the material is below the mentioned specific activity level.
Measurements of Production and Distribution of Radionuclides in the Spallation Target
Article CITATION 1 READS 5 4 authors, including: Some of the authors of this publication are also working on these related projects: Modeling of nuclear reactions for intermediate and high energies induced by protons, deuterons and carbon ions. Modeling of generation of radioactive isotopes. View project ABSTRACT Designing of an Accelerator Driven System needs a thorough evaluation of build-up of long-lived radioactivity and changes in elemental composition in construction materials. They can be calculated with the use of experimentally verified computer codes and data libraries. The experiment presented here is thought as a part of such verification. Samples of Pb and Bi were placed respectively in and behind a Pb target irradiated with 650 MeV protons. Gamma spectra of samples were counted and analysed. Activities of several radionuclides were determined and some of them compared with values obtained from MCNPX calculations.
Leaching of Accelerator-Produced Radionuclides
Health physics, 1997
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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.
2007
We have written and tested a scripting tool which extracts cell-based material, neutron fluxes and spallation product information from MCNPX output files and drives transmutation calculations using the CINDER'90 code in accelerator-driven systems. This tool was later extended to also drive the alternative transmutation codes ORIHET3 and SP-FISPACT. Another scripting tool was generated for extracting decay gamma spectra and preparing MCNP(X) source decks for remnant decay gamma field calculations. The emphasis of this development effort was put on minimizing the need for user supplied input and simplifying the process of complex activation analyses. Also, an isotope production tally was implemented into MCNPX to enable feeding the transmutation codes without having to post-process the MCNPX history tape with HTAPE3.
1990
This study presents a basic methodology for characterizing low-level radioactive waste. The methodology takes into account the source of the waste, the characteristics of the waste, the potential for the presence of specific radionuclides, the characteristics associated with each radionuclide, and the applications of and limits on the techniques used to measure radiation levels. Waste generators have shown an interest in such a methodology, based on the increased regulatory requirements for handling, transporting, and disposing of this waste. One reason no formal methodology has yet been implemented is that limits have not been set to define the point at which a waste material is considered to be low-level and below which it is "below regulatory concern" (BRC) or contains no more than a diminimus level of radioactivity (i.e., is nonradioactive). This study introduces the concept of diminimus limits specific to waste streams and waste forms. It also provides a mechanism for evaluating waste in relation to BRC limits. The methodology employs both standard and theoretical assessment, analysis, and data interpretation techniques. The computer code Microshield is used to simulate the decay of radionuclides and radiation levels of low-level waste in a theoretical study. The methodology developed here can be used to perform a iii T-4020 complete and adequate characterization of waste and as a sound basis for determining and documenting when a waste is BRC or when it contains no more than a diminimus level of radioactivity. The methodology ultimately relies on the use of limits for classification; if these limits cannot be established, the waste is conservatively handled as low-level. iv T-4020
Application of accelerators in nuclear waste management
A key roadblock to development of additional nuclear power capacity is the concern over management of nuclear waste. Nuclear waste is predominantly comprised of "spent" fuel discharged from operating nuclear reactors. The 104 operating US light water reactors (LWRs), that currently produce about 20% of the US electricity or more than 70% of the U.S. emission-free electricity, and, given the life extension of present plants, will create about 120,000 tons of such spent fuel over the course of their lifetimes. Worldwide, more than 250,000 tons of spent fuel from reactors currently operating will require disposal. The toxicity of the spent fuel, mainly due to ionizing radiation, will affect future generations for long into the future. The large quantity and its long-lived toxicity present significant challenges in waste management.