Gamma Energy Deposition Research Articles

Implementation of a machine learning technique for estimating gamma direction using a coaxial High Purity Germanium detector

We demonstrate the ability to obtain the direction of the gamma rays using a standard coaxial high purity germanium (HPGe) detector using the direction-sensitive information embedded in the shape of the pre-amplified HPGe signals. We deduced the complex relationship between the shape of the signal and the direction from which the gamma-ray enters the detector active volume using a two-step machine learning technique. In the first step, we collected pulses from the HPGe detector due to a 133Ba radioactive source placed in four distinct positions around the detector while keeping the distance from the center of the detector crystal constant. A subset of the pulses collected with radioactive source kept at the four positions was used to train an artificial neural network (ANN) called a self-organizing map (SOM) to cluster the HPGe waveforms based on their shape. The trained SOM network was then utilized to produce direction-specific maps corresponding to pulses generated when the 133Ba source is at a specific location with respect to the detector. In the second step, we used the SOM-generated direction-specific maps to train another network composed of a single feedforward layer for predicting the direction of the gamma ray from the pulses produced by the HPGe detector due to the gamma energy deposition. Our results show that even without employing complex methodologies, a standard coaxial HPGe detector can estimate the direction of incoming gamma rays and thus, provide initial guidance on the gamma-emitting radioactive source direction with reference to the detector.

The gamma deposition matrix method for gamma heating calculations for LWRs

This paper presents a new method called the gamma deposition matrix (GDM) method to calculate gamma energy deposition without performing a full-core explicit gamma transport calculation. This method computes gamma energy deposition by using the known gamma source and the GDM to perform a simple matrix-vector multiplication. The entries of the GDM represent the gamma energy deposited in a given spatial cell due to a gamma source in another (or the same) spatial cell. Since microscopic cross sections of gamma interactions are insensitive to temperature, and the macroscopic cross sections are insensitive to isotope depletion, the GDM can be pre-calculated and stored in advance and used in the coupled (n,γ) calculations with thermal-hydraulic feedback and core depletion. Three approximations are made to reduce the size and generation time of the GDM: 1) eliminating the dependence of the GDM on gamma energy, 2) neglecting the GDM for gamma deposition far from the gamma source and 3) approximating the GDM for gamma deposition out of a certain range but not too far from the gamma source. Since the second approximation may result in a loss of gamma energy, the third approximation includes a correction technique to conserve the overall gamma energy deposition.The method and all these approximations are proved to be accurate and effective for reducing the memory and runtime for realistic 2D whole-core cases. For a 2D small modular reactor whole core case with 37 assemblies and 289 pins within each assembly, the gamma heating calculation is accelerated by a factor of 3.4 while maintaining satisfactory accuracy.

On-line measurement and simulation of the in-core gamma energy deposition in the McMaster nuclear reactor

In a nuclear reactor, gamma radiation is the dominant energy deposition in non-fuel regions. Heat is generated upon gamma deposition and consequently affects the mechanical and thermal structure of the material. Therefore, the safety of samples should be carefully considered so that their integrity and quality can be retained. To evaluate relevant parameters, an in-core gamma thermometer (GT) was used to measure gamma heating (GH) throughout the operation of the McMaster nuclear reactor (MNR) at four irradiation sites. Additionally, a Monte Carlo reactor physics code (Serpent-2) was utilized to model the MNR with the GT located in the same irradiation sites used in the measurement to verify its predictions against measured GH. This research aids in the development of modeling, calculation, and prediction of the GH utilizing Serpent-2 as well as implementing a new GH measurement at the MNR core. After all uncertainties were quantified for both approaches, comparable GH profiles were observed between the measurements and calculations. In addition, the GH values found in the four sites represent a strong level of radiation based on the distance of the sample from the core. In this study, the maximum and minimum GH values were found at 0.32 ± 0.05 W/g and 0.15 ± 0.02 W/g, respectively, corresponding to 320 Sv/s and 150 Sv/s. These values are crucial to be considered whenever sample is planned to be irradiated inside the MNR core.

Tungsten boride shields in a spherical tokamak fusion power plant

The favourable properties of tungsten borides for shielding the central high temperature superconductor (HTS) core of a spherical tokamak fusion power plant are modelled using the MCNP code. The objectives are to minimize the power deposition into the cooled HTS core, and to keep HTS radiation damage to acceptable levels by limiting the neutron and gamma fluxes. The shield materials compared are W2B, WB, W2B5 and WB4 along with a reactively sintered boride B0.329C0.074Cr0.024Fe0.274W0.299, monolithic W and WC. Five shield thicknesses between 253 and 670 mm were considered, corresponding to plasma major radii between 1400 and 2200 mm. W2B5 gave the most favourable results with a factor of ∼10 or greater reduction in neutron flux and gamma energy deposition as compared to monolithic W. These results are compared with layered water-cooled shields, giving the result that the monolithic shields, with moderating boron, gave comparable neutron flux and power deposition, and (in the case of W2B5) even better performance. Good performance without water-coolant has advantages from a reactor safety perspective due to the risks associated with radio-activation of oxygen. 10B isotope concentrations between 0% and 100% are considered for the boride shields. The naturally occurring 20% fraction gave much lower energy depositions than the 0% fraction, but the improvement largely saturated beyond 40%. Thermophysical properties of the candidate materials are discussed, in particular the thermal strain. To our knowledge, the performance of W2B5 is unrivalled by other monolithic shielding materials. This is partly as its trigonal crystal structure gives it higher atomic density compared with other borides. It is also suggested that its high performance depends on it having just high enough 10B content to maintain a constant neutron energy spectrum across the shield.

Delayed gamma determination at the JSI TRIGA reactor by synchronous measurements with fission and ionization chambers

Well characterized neutron and gamma fields inside a nuclear reactor are of key importance for its safe operation and for successful utilization of various research reactor irradiation facilities. In case of high-flux research reactors such as BR2 in Belgium, Maria in Poland and the future Jules Horowitz Reactor in France, the gamma energy deposition rate in reactor structural components and irradiated samples reaches values well over 10W/g. To assure safe reactor operation, the gamma field and associated heating must therefore be thoroughly characterized in order to provide adequate component and sample cooling.The gamma field can be divided into two contributions: prompt gamma rays are emitted almost instantly after neutron interaction with nuclei, while the delayed gamma rays are emitted from nuclei, which become radioactive by neutron absorption, generated from fission and other processes. Most modern Monte-Carlo particle transport codes enable the transport of prompt gamma rays; a few support delayed gamma ray generation and transport as well. The latter have mostly been applied to fusion devices, where detailed shutdown dose-rate measurements have been performed. Although the delayed gamma field can also be simulated in fission devices, significant inaccuracy in the result is to be expected due to the computational complexity arising from the large number of radioactive fission products and incompleteness of nuclear data. Furthermore, the unavailability of experimental delayed gamma measurements in fission systems presents an important challenge for the validation of the experimental results.Previous measurements in several research reactors show that the delayed gamma flux amounts to around 30% of the total gamma flux. However, these evaluations were performed with measurement data obtained during rapid reactor shutdowns (SCRAMs), using a single measurement point per SCRAM.In this paper we propose a new technique to accurately determine the magnitude of the delayed gamma component and its time evolution, based on synchronous acquisition of fission and ionization chamber signals. The measurements were performed at the JSI TRIGA reactor, using fission and ionization chambers placed in several in-core measurement positions. Their signal was acquired synchronously and at the highest possible acquisition rate in order to distinguish between measurement noise and reactor transients. Using the novel delayed gamma extraction technique we were able to estimate the magnitude of the delayed gamma contribution to be: 18.9% ± 2.0% at the reactor core periphery, linearly increasing towards the reactor core center to 31.4% ± 2.8% of the total measured gamma flux signal after 10min of reactor operation.

Improving sensitivity of a BEGe-based high-purity germanium spectrometer through pulse shape analysis

We performed Pulse Shape Analysis to separate single-scattered gamma energy deposition events from multiple-scattered photons in a high-sensitivity gamma -ray spectrometer. The spectrometer is based on a Broad Energy High Purity Germanium detector and the developed technique uses multivariate analysis by an application of the Multi-Layer Perceptron Neural Network. A very good separation of the single-site- and multi-site events was achieved leading to a significant reduction of the background level of the investigated spectrometer – the double escape peak, rich in single-site events, was reduced by 95%, while the full energy peaks lost at most 25% of their counts. The peak to Compton ratio, calculated for the 2614.5 keV gamma line from ^{208}Tl, was improved by 114.3%.

Gamma transport and diffusion calculation capability coupled with neutron transport simulation in KARMA 1.2

KAERI has developed a lattice transport calculation code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and its library generation system. Recently, the library generation system has been extended to include a gamma cross section generation capability and a gamma transport/diffusion calculation module has been implemented in KARMA 1.2. The method of characteristics for the neutron transport calculation to estimate eigenvalue has been utilized to predict gamma flux distribution and energy deposition. In addition, the coarse mesh finite difference method with diffusion approximation has also been utilized to estimate gamma flux distribution and energy depositions for each coarse mesh with homogenized pins as a computationally efficient alternative. This paper describes the procedure to generate neutron induced gamma production and gamma cross section data, and the methods to predict gamma flux distribution, gamma energy deposition and gamma smeared pin power distribution. The computational results for benchmark problems show that the gamma library and gamma simulation in KARMA are reasonable. And it is noted that gamma smeared power distributions predicted by coarse mesh diffusion calculation are very accurate compared to the results of transport calculation.

Analysis on nuclear heating in the superconducting coils of HT-7U tokamak fusion device

The HT-7U superconducting tokamak will generate neutrons at a rate of the order of 10 15 per second during the D–D operation with long pulse up to 1000 s. Nuclear heat in the Toroidal Field (TF) and Poloidal Field (PF) coils have been estimated by using the Monte Carlo particle transport code MCNP-4C and the latest version of the Fusion Evaluated Nuclear Data Library (FENDL-2) based on the three-dimensional geometrical configuration. The calculations show that the nuclear heat in the coils originates ∼10% from neutron energy deposition and ∼90% from induced gamma energy deposition. To reduce the nuclear heat in the superconducting magnets of the TF and PF coils, a double-wall-structured vacuum vessel (VV) with the boronated water layer of thickness of 5 cm is designed to shield against neutron irradiation.

Coolant temperature trends in EBR-II reflector subassembliest

The radial-reflector subassemblies of the Experimental Breeder Reactor-II (EBR-II) are in the form of large hexagonal-shaped steel blocks inserted into a hexagonal-shaped steel duct with sodium coolant flowing upward in the thin space between the blocks and duct. This geometry combined with the small coolant flow rates and gamma energy deposition rates cause the coolant temperature distributions to be greatly affected by intersubassembly heat transfers. Reported are calculated trends of the coolant temperatures in the six facial intraduct gaps of radial reflector subassemblies surrounding a seven-rowed core and a six-rowed core. Effects of overall changes of reflector region flows and powers are also examined.

Gamma heating measurements with proportional counters

A new data acquisition technique (the continuously-varied bias-voltage acquisition mode) has been developed and tested for the low-flux broad-energy regime characteristic of existing fusion blanket mockups. This method of analysis allows for the acquisition of data spanning several orders of magnitude in energy with a single proportional counter. Utilizing this method, the gamma energy deposition in a mixed neutron and gamma field was measured.

An Analysis of Gamma-Ray Energy Deposition in a Heterogeneous Liquid-Metal Fast Breeder Reactor

Gamma-dose analysis was performed on an absolute-value basis at the zero-power plutonium reactor (ZPPR)-13B/4 critical assembly, which was one of the benchmark radially heterogeneous 650-MW (electric) liquid-metal reactor cores, to validate the current data and methods applicable to power reactor design calculations. Discussions of particular aspects inherent to the critical measurements, such as heterogeneity in the ZPPR plate cells and delayed fission product gamma effect, are included. The results show that the gamma-energy deposition distribution was reproduced within an accuracy of {approximately} 10% in cores that had complicated fuel and internal blanket layouts.

Scaling of Gamma Dose Rate Upset Threshold in High Density Memories

Pulsed gamma upset threshold for high density RAMs is explored using constant field scaling rules. The analysis indicates that the upset threshold for DRAMs will not decrease with increased levels of integration and may improve relative to present technology. Potential improvement may be limited by statistical effects in gamma energy deposition. It is shown that statistical effects are magnified both by the increasing number of memory cells on a chip and by their extremely small feature size. Extreme value statistical techniques are used to determine worst-case cell memory loss. The paper also presents parallels of pulsed gamma and single event upset (SEU) effects that may be useful in SEU hardness assurance for VLSI memories.