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Jan 29, 2022, 9:16:16 AM
Jan 29, 2022, 11:16:16 PM
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An in-situ alpha air monitor for the retrieval of fuel debris at the Fukushima Daiichi Nuclear Power Station (#108)
F. Honda1, Y. Tsubota1, Y. Tamakuma2, S. Tokonami2, A. I. Ohno1, T. Nakagawa1
1 Japan Atomic Energy Agency, Ibaraki, Japan
A large number of radioactive aerosols, especially alpha particles, are expected to be generated during the fuel debris retrieval on the Fukushima Daiichi Nuclear Power Station. From the viewpoint of preventing contamination of the surrounding environment and the vicinity of the human-access area, it is important to measure the activity concentration of airborne radioactive substances inside the primary containment vessels (PCVs). In particular, it is necessary to monitor the concentration of particulates containing α-nuclides (α-aerosols), which have high effective dose coefficients upon inhalation. This presentation reports the development of an in-situ alpha air monitor (IAAM) for direct measurement of α-aerosols by combining a flat-type flow path (FFP), an air heater, a ZnS scintillator, and a multi anode photomultiplier tube. The monitor should operate under high humidity with the maximum counting rate of approx. 2.1×107 cpm. To achieve the two requirements, the monitor was designed to keep the air sufficiently dry without overheating the detector, and to reduce the detection of coarse particles. This study also conducted a basic performance test using the developed IAAM with a modified FFP. As a result, we could keep the humidity of the air less than 30 %RH by heating the inlet of the FFP to 80 °C. In addition, by placing the FFP in a vertical position and installing a bend at the air intake port, coarse particles were reduced approx. 1/2-1/3. These achievements enable the monitor to measure α-aerosols more precisely in the viewpoint of internal exposure assessment.
Keywords: Radiation monitoring, Alpha particles, Nuclear fuels
Improvement of Clearance Reliability for Plastic and Metal Mixture Waste Using CT (#160)
H. Tagawa1, J. Kawarabayashi2, T. Yoshii3, 1, N. Hagura2
1 Tokyo City University, Cooperative Major In Nuclear Energy, Setagaya, Japan
Clearance is the removal of radioactive material or radioactive objects from regulatory control by a regulatory body. One method of evaluating the radioactivity for clearance is measuring the storage container containing the waste from the outside with a radiation detector. Formerly, concrete and metal were the only substances that could be released, if the object consisted of one element, by measuring by this method in Japan. Due to the amendment of the clearance regulations in 2020, plastic and metal mixtures (power cables, switchboards, etc.) can be released under the clearance. However, when the storage container containing these substances is measured as it is, the non-uniformity of the density of substances in the storage container makes it difficult to consider the actual radiation shielding effect, and it must be evaluated extremely conservatively. In order to incorporate a reliable radiation shielding effect into radionuclide concentration evaluation, we propose to introduce a non-destructive method, such as a CT scan that can be used to measure the elements in the container from the outside. CT-scans can only provide values proportional to the attenuation coefficient of X-rays which is insufficient to contribute to the evaluation of the shielding effect for the radiation emitted from radioactive material. Therefore, assuming that the elements contained in the storage container are known as prior information, we experimentally verified whether it is possible to discern the elements by CT scan. In this study, rods of PMMA, aluminum, copper, and iron in a small container were scanned by CT and the images were reconstructed by FPB and ML-EM methods. The results showed that each element could be distinguished from the others successfully when the voltage of the X-ray tube was over 100 kV.
Keywords: Clearance, CT, decommissioning
An advanced, blind-tube monitoring instrument to characterize subsurface radioactive plumes (#401)
S. Elisio1, J. Graham2, M. Joyce1
1 Lancaster University, Engineering Department, Lancaster, United Kingdom
This research presents a design of a resilient radiometric logging probe, which satisfies key, operational site constraints, for direct characterization of underground radioactivity, specially caesium-137 and strontium-90, at interim storage sites for radioactive waste. The probe comprises a commercially-available Ø10mm×9.5 mm CeBr3 scintillator detector attached to compact digitizer unit, in a resistant and waterproof housing. The probe is designed to be lowered down into metallic blind tubes by means of a winch system. The preliminary detector tests are reported in this paper such as spectral response and angular dependence to gamma radiation. Modelling and laboratory experiments are underway to validate the concept and calibrate the system in a soil-filled phantom arrangement that replicates the in-ground blind-tubes set on site which are also reported on in this research.
Keywords: blind-tube logging, CeBr3 scintillator, gamma-ray spectroscopy, decommissioning
Definition of a new differential calorimeter assembly CALORRE for the measurement of the nuclear heating rate inside the MIT reactor (#448)
A. Volte1, M. Carette1, A. Lyoussi2, G. Kohse3, C. Reynard-Carette1
1 Aix Marseille Univ, Université de Toulon, CNRS, IM2NP, Marseille, France
The project leading to this publication has received funding from the Excellence Initiative of Aix-Marseille University - A*Midex, a French “Investissements d’Avenir” programme.
This paper deals with the study of a new differential calorimeter assembly CALORRE. This assembly is intended to measure nuclear heating rate inside an in-core water-loop in the MIT reactor (MITR). It is obtained from two new cells. Their configuration design has been modified. A first change is done to eliminate the thermal contact resistances observed with previous prototypes and having an influence on the response of the calorimeter. The vertical fin of each calorimetric cell has therefore been increased to act as a jacket and the sample is machined as a single block in the cell head. The second change allows the adjustment of the calorimeter sensitivity due to the expected nuclear heating rate (2W.g-1 in stainless steel). The last change concerns the cell size and the assembly method to propose a low sensor height. One of the major concerns of this paper is to interpret the numerical response of the new calorimeter under real conditions obtained thanks to a previously validated 3-D thermal model for an adapted nuclear heating range. Finally, the linearity of the response, the sensitivity, the maximal and wall temperatures and the time response are studied and will be compared to previous experimental responses of similar CALORRE cells configuration (nature of structure material and vertical fin configuration).
Keywords: Calorimeter, Nuclear Heating Rate, On-line Measurements, Calibration, Irradiation campaign
Optimizing plastic scintillator detector geometries to improve γ-discrimination in mixed n/γ fields (#948)
A. J. Parker1, M. D. Aspinall1
1 Lancaster University, Engineering Department, Lancaster, United Kingdom
The performance of varying conical and cylindrical plastic scintillator geometries was simulated to provide improved pulse shape discrimination (PSD) in mixed n/γ fields. Using GEANT4, planar 0.1-10 MeV mixed n/γ fields were simulated interacting with EJ-276 plastic scintillator detectors in 52 conical and cylindrical geometries. A planar mixed n/γ source was positioned 20 cm from the centre of the detectors and rotated around 90° in the plane of the detector height. Using these simulations, the ratio of the total energy deposited by neutrons and γ-rays was determined, as was the ratio of the optical photons hitting the photocathode produced by both radiation types for the different detector geometries. Results show conical geometries can provide up to a 10% increase in the ratio energy deposited by neutrons to that of γ-rays. Additionally, a higher percentage of optical photons produced by neutrons are detected in the photocathode. However, the overall detection efficiency decreases. This work indicates that relatively significant increases in PSD performance could be attained by simply changing the detector shape.
AcknowledgmentThis work was funded by the United Kingdom's Engineering and Physical Sciences Research Council, through grant EP/T013532/1, as part of the EPSRC UK-Japan Civil Nuclear Research Programme to develop radiation tolerant criticality monitoring instrumentation.
Keywords: Detector optimization, fast neutron, mixed field analysis, pulse shape discrimination
Miniature Germanium Detector For Field Gamma-ray Operation (#1059)
M. Ginsz1, D. Ralet1, V. Marian1, P. Quirin1
1 Mirion Technology (Canberra) SAS, Lingolsheim, France
We developped a new miniature High Purity Germanium detector designed for use in high gamma-ray flux environment, very tight spaces, as well as for field applications such as nuclear material characterization and decommissioning. The lightweight design (1.8 kg) and low power consumption (~12W) makes it possible to embark the detector onto remote-controlled robotic platforms (terrestrial or aerial) in order to perform high resolution radiation measurements in highly contaminated areas. The γ-ray spectrometed exhibits excellent spectroscopic performance (≤ 1.0 keV@122 keV, ≤ 1.6 keV@662 keV, ≤ 2.2 keV@1.33 MeV) and has almost immediate availability (cool-down time below 30 minutes), making it an ideal alternative to Room Temperature Semiconductor Detector (RTSD) solutions having with inferior spectroscopic and count-rate capabilities.
Keywords: HPGe, cryocooler, gamma-ray spectroscopy, high flux, handheld