Research Reactor Research Articles

Testing of an organic metal halide perovskite for fast neutron detection

In this work, we synthesized and characterized the Methylhydrazinium Lead Trichloride MHyPbCl3(CH3NH2NH2PbCl3) perovskite as a fast neutron detector. The high hydrogen density of MHyPbCl3 enables efficient energy conversion from a fast neutron into a recoiled proton through the 1H(n,n)1H elastic scattering interaction, thereby, allowing for direct charge detection. Through IV characterization and X-Ray excitation, the crystal demonstrated a high resistivity at 4.43E11Ωcm and a good mobility-lifetime product (μτ) of 9.1E-3 cm2V, respectively, under C60/BCP/Cu and Au contact configuration to form an ohmic-ohmic detector. The crystal showed a good sensitivity to X-rays using an x-ray tube. The feasibility of the direct neutron conversion detector is demonstrated using the fast neutron beam at a Research Reactor. Waveforms from a charge-sensitive pre-amplifier showed distinct radiation-induced pulses from the MHyPbCl3 detector in response to the reactor neutron beam. Using a thermal neutron filter and gamma shielding in the beam, we showed that the pulses produced were more likely from neutron interactions, despite the Pb containing MHyPbCl3 is also sensitive to gamma-rays. With those fast neutron pulses, a post-pulse processing code was used to conduct pulse height analysis (PHA) and reconstruct an energy spectrum.

Gamma noise to non-invasively monitor nuclear research reactors

Autonomous nuclear reactor monitoring is a key aspect of the International Atomic Energy Agency’s strategy to ensure nonproliferation treaty compliance. From the rise of small modular reactor technology, decentralized nuclear reactor fleets may strain the capacities of such monitoring and requires new approaches. We demonstrate the superior capabilities of a gamma detection system to monitor the criticality of a zero power nuclear reactor from beyond typical vessel boundaries, offering a powerful alternative to neutron-based systems by providing direct information on fission chain propagation. Using the case example of the research reactor CROCUS, we demonstrate how two bismuth germanate scintillators placed outside the reactor vessel can precisely observe reactor criticality using so called noise methods and provide core status information in seconds. Our results indicate a wide range of applications due to the newly gained geometric flexibility that could find use in fields beyond nuclear safety.

Delayed Heating Calculations Using the MCNP-ORIGEN Activation Automation Tool

An accurate assessment of the deposited energy across a reactor geometry allows for a better determination of the heat removal requirements and ensures effective cooling after shutdown. This work discusses several methods in detail that target the heat deposition in specific reactor regions, leading to a choice of a prevalent method for the calculation workflow. This paper introduces a delayed heating calculation workflow based on the formal three-step process. The workflow relies on the MCNP-ORIGEN Activation Automation tool for performing the first two steps of the process, while the third step is conducted via MCNP photon transport simulations. This paper showcases two applications to demonstrate the workflow and simulation outputs. These include an Advanced Test Reactor (ATR) experiment and the RA-6 reactor structures.

Study on producing radioisotopes based on fission or radiative capture method in a high flux reactor

Radioisotopes tend to play important roles in many fields, such as industry, healthcare, agriculture, aerospace, etc. Radioisotope production is mainly through accelerators or research reactors, and high flux research reactor is one of the most effective approaches for radioisotope production. The physical basis of preparing radioisotope relies on nuclear reactions occurring in the reactor core, which includes fission, (n,γ), (n,α), and (n,p) reaction, etc. Among them, fission and (n,γ) reaction are most important in the nuclear reactor. For example, the 99Mo could be generated by uranium fission and extracting from the fission products, or through the radiative capture reaction from enriched 98Mo. As for the fission method, the irradiation target is gradually transitioning from high enriched uranium (HEU) target to low enriched uranium (LEU) target due to the requirement of non-proliferation. In this paper, studies on the impacts of different fission targets on radioisotope productions are conducted. Moreover, an optimized study on the radiative capture method is performed to improve the production efficiency. It is concluded that it is advantageous to use radiative capture method to generate radioisotopes in high flux reactor, which helps to improve the specific activity with environmental friendliness.

The Application of Laser-Scanning 3D Model Reconstruction Technology for Visualizing a Decommissioning Model of the Heavy Water Research Reactor

Currently, over 100 nuclear power units globally have been in operation for more than 40 years. Hindered by the limitations of computer technology at the time, these nuclear facilities lack detailed electronic drawings. Activities such as equipment replacement and process circuit system modifications during operation result in discrepancies between paper drawings and actual conditions. Given the complexity and irreversibility of nuclear facility decommissioning activities, virtual simulation technology is often employed before the decommissioning process begins to assist in designing and validating decommissioning plans. Consequently, the creation of high-precision 3D models is crucial for subsequent decommissioning designs. Through innovatively utilizing laser-scanning 3D model reconstruction technology in the reconstruction of the model of China’s first heavy water research reactor undergoing decommissioning, this paper provides an overview of the process of laser-scanning 3D model reconstruction and its application in reconstructing the heavy water research reactor model. Using a 3D laser scanner, four decommissioning areas of the heavy water research reactor, including the reactor building, secondary water pump room, ventilation center, and low-level radioactive wastewater storage tank area, were subjected to 3D laser scanning. The acquired point cloud data from 572 scanning stations were processed using point cloud processing software for denoising, stitching, and triangulation. The triangulated model was then imported into modeling software for 3D reconstruction, ultimately establishing a digitalized model of the heavy water research reactor suitable for subsequent decommissioning simulation and design.

Assessing the PLEIADES/GERMINAL V3 fuel performance code against transmutation experiments in sodium fast reactors

We present the results of a first assessment of the multiphysics PLEIADES/GERMINAL V3 fuel performance code against integral irradiation experiments of americium-bearing MOX fuel pins irradiated in sodium fast reactors. The pins considered in this work cover a range of americium contents from 2 to 5 weight %, being representative for the homogeneous transmutation path. The code predictions are compared to several post-irradiation examination results on integral (e.g., fission gas release) and local quantities (e.g., actinides redistribution in the fuel pellet), and demonstrate a satisfactory agreement with experimental data. The deviations between calculated and experimental quantities shed some light on modeling developments needed to better calculate the behavior of americium-bearing mixed-oxides fuel pins under irradiation.

PCMI Uncertainty and Sensitivity Analysis of Es-Salam Reactor Fuel Rods During Steady-State Operations

Nuclear safety relies heavily on the quality of the results of numerical simulation codes. Among the various components of the simulation of the installation are the pellet-cladding mechanical interaction (PCMI), and the peak cladding temperature (PCT). Although the correlations describing the physical, mechanical, chemical, and thermal phenomena that occur in nuclear installations have reached a high level of quality, there remain uncertainties on the final results due to uncertainties in the input parameters which cannot be eliminated. A realistic estimate of these uncertainties is necessary to evaluate the reliability of the simulation results. When the best-estimate approach plus uncertainty (BEPU) is employed in the design of a nuclear installation, design-basis accidents are studied more realistically. This method must be used even in the design of research reactors because they are at the origin of any development of nuclear technology. We propose through this study an uncertainty and sensitivity analysis of PCMI and PCT of a heavy water nuclear research reactor fuel rod. To determine the input parameters that influence PCMI and PCT, we utilize the FEMAXI-6 code. The thermodynamic table of the FEMAXI-6 code is adapted to the case of heavy water. Two system codes are used for uncertainty and sensitivity analysis: RELAP5 and PARET. The study confirmed that in the event of a shortage of heavy water, light water can be injected in its place to remove decay heat from the core and shut down the reactor safely. The safety margin between the PCT and the saturation temperature is reduced from about 10°C in the conservative approach to less than 1°C by the BEPU approach.

In-situ removal of microcystin aeruginosa and microcystin-LR by biochar supported sulfide nZVI via persulfate activation: Performance, mechanism and degradation pathway

As a hepatotoxin generated by the excessive proliferation of Microcystis aeruginosa, Microcystin-LR (MC-LR) has posed serious threat to humans and wildlife. In this study, biochar supported sulfide nZVI and persulfate sustained-release agent (S-nZVI-BC/PSSA) were prepared to simultaneously remove MC-LR and Microcystis aeruginosa. Batch experiments showed that the removal efficiency of MC-LR (100 μg/L) by S-nZVI-BC/PSSA could reach 92.0 % when the BC pyrolysis temperature, the Fe/C mass ratio and the Fe/S molar ratio were 700 °C, 1:2 and 40:1, respectively. According to the results of EPR and density functional theory (DFT), the carbon–carbon conjugated diene bond (C = C-C = C) on Adda were most vulnerable to attack by the free radicals such as •OH and SO4-• generated by S-nZVI-BC/PSSA. The results of reactor experiments indicated that S-nZVI-BC/PSSA had good removal effect on MC-LR and Microcystis aeruginosa. According to the transcriptome analysis, S-nZVI-BC/PSSA led to the disorder of cell metabolic process by up-regulating ATP and pigment synthesis genes and down-regulating photosynthesis genes, eventually inhibiting the growth of Microcystis aeruginosa and even causing their death. In addition, the up-regulation of biological transport processes could enhance the release of MC-LR from Microcystis aeruginosa, which was completely removed by S-nZVI-BC/PSSA. Moreover, four degradation pathways of MC-LR were analyzed, and the final products were CO2 and H2O. Therefore, this S-nZVI-BC/PSSA system showed great promise for practical applications in the removal of Microcystis aeruginosa and MC-LR.

Preparation for dismantling the liquid waste disposal system of MR and RFT research reactors

Preparation for dismantling the liquid waste disposal system of MR and RFT research reactors

Impact of differences between the Nuclear Law Act and the Vienna Convention on Civil Liability for Nuclear Damage on the Terms and Conditions of Nuclear Civil Liability Insurance

The focus of this article is that Polish legal provisions on civil liability for nuclear damage do not reflect the provisions of the Vienna Convention in all aspects. However, due to some limitations hereof, the considerations presented deal only with some selected issues that may affect the availability of nuclear third party liability insurance, and will be divided into two parts.The aim of the first part is to consider the major differences between the provisions of the Nuclear Law Act and the Vienna Convention, which include, among others, different deadlines for seeking compensation for nuclear personal injury and/or methods for satisfying such claims. Since both issues refer to personal injury, the notion of the latter is analysed, bearing in mind that certain discrepancies between the Polish Act and the Vienna Convention can also be noted.The second part discusses terms and conditions of third party nuclear liability insurance, entities specializing in offering such insurance, including nuclear pools and nuclear mutual insurance companies, as well as obstacles that may prevent third party nuclear liability insurance from being offered on the Polish market as compulsory insurance.This article covers only the rules relating to civil liability for nuclear personal injury resulting from the operation of a nuclear power plant, with those rules being considered primarily from the perspective of the insurance practice. The principles of civil liability for the transport of nuclear material from nuclear facilities, the operation of research reactors or of radioactive waste storage facilities have not been taken into account, as both insurers’ approach and terms and conditions of insurance covering said activities may vary slightly when compared with nuclear power plants.

JSI TRIGA fuel rod reactivity worth experiments for validation of Serpent-2 and RAPID fuel burnup calculations

Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient ΔρBU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.

Temperature feedback reactivity analysis for LEU-Fuelled SLOWPOKE-2 research reactor using DRAGON5 and DONJON5 codes

The purpose of this study is to provide a new deterministic model for the SLOWPOKE-2 nuclear research reactor at École Polytechnique de Montréal (EPM) with LEU (Low Enriched Uranium) core. Using the latest release of the code system DRAGON5 and DONJON5 and the cross-section data library ENDFB.VII rel.1 evaluation, the developed model is applied to simulate the neutronic behavior of the SLOWPOKE-2 research reactor. We studied the separate temperature effects of the main components of the core (i.e., fuel, coolant/moderator, beryllium reflector, and water reflector). The contribution of different physical phenomena to the RTC was assessed. The temperature reactivity feedback calculated using the deterministic approach based on the DRAGON5 and DONJON5 code system using the ENDF/B-VII.1 evaluated nuclear data library produced in the WIMD-D4 format is in good agreement. Therefore, this work proves the capability of DRAGON5 and DONJON5 codes, normally used for power reactors, to reliably simulate a low-power research reactor.

Reproducible Benchmark for the SNAP 8 Experimental Reactor at Dry Conditions

This work provides fully reproducible benchmark models of the Systems for Nuclear Auxiliary Power (SNAP) 8 Experimental Reactor. Validation benchmarks of the criticality configuration experiments under dry conditions with no coolant are presented. In addition, relevant reactivity worth experiments, and measurement of power distributions are validated and presented here. Discrepancies between modeled and experimental results are at most 110 pcm for fuel and poison worths, and critical configurations that do not manipulate control elements are within 50 pcm. Larger discrepancies found in reflector element worth experiments are due to experimental methodology which was noted as being simplified due to limited calculational capabilities at the time, while discrepancy in the control element manipulated critical configuration is due the absence of structural components for modeling simplification. Power distributions closely follow what was observed in experiment with expected peaking in reflecting elements. All models are well documented with cited references describing and justifying assumptions, simplifications, and adjustments to reproduce the models with any nuclear safety codes; model inputs and outputs are stored in the snapReactors GitHub repository.

Research on sensor data optimization technology for thermal hydraulic experiment of nuclear reactor

The sensor signals collected by the nuclear reactor thermal hydraulic experimental system are mixed with complex noise information. The uncertainty of sensor data directly affects the analysis and evaluation effect of machine learning algorithms on the operating status of the experimental system. Traditional denoising methods have poor adaptability to different types of noisy data, are highly dependent on researchers and have high design cost. This paper proposes a data optimization technology based on a hybrid adaptive real-time denoising (HART) model, which can realize data distribution self-adaptation and algorithm hyperparameter self-optimization according to the characteristics of noisy source data. Through the joint denoising algorithm with Variational Mode Decomposition (VMD) algorithm as the core, the partition denoising of source data is realized. In addition, the practical application of the source data of thermal hydraulic experiments in the nuclear field and the verification of simulation and noise addition experiments have been completed. The results show that the denoising optimization model proposed in this paper has the characteristics of good adaptability, self-optimization and real-time processing for different sensor signals. It can effectively remove noise information, restrain data uncertainty, and provide high-quality data source for experimental data analysis based on machine learning. At the same time, the optimization technology can be further applied to the optimization of sensor data of nuclear power plants (NPP).

Возможности создания источников излучения для персонализированной брахитерапии на основе 3D-каркасов из сплава титана

Purpose: The study explores the possibility of manufacturing radiation sources for personalized brachytherapy using titanium alloys, activated in a neutron flux reactor, by measuring the radiation composition of applicator implants and their dosimetric characteristics. Material and methods: A 3D implant of a brachytherapy source was made from a titanium alloy using an additive selective laser melting setup. The titanium 3D prototype was irradiated for three days in the horizontal experimental channel of the IR-8 reactor. Subsequently, measurements of the gamma-ray spectrum from the irradiated implant were carried out on a spectrometer, and dose characteristics of the 3D implant were measured using a dosimeter-radiometer. Results: In the experimental 3D implant obtained by us, the radionuclide 47Sc exhibits the highest activity. Currently, 47Sc is considered a promising candidate for brachytherapy. It possesses attractive nuclear and physical properties as a β-emitter, decaying into the ground state (27 %) of 47Ti (Eβmax = 600 keV) and the excited state of 47Ti (Eβmax = 439 keV) with a half-life of 3.4 days. Additionally, 47Sc emits γ-radiation at an energy of 159 keV (68 %), which is suitable for imaging, allowing for SPECT or planar scintigraphy and obtaining a picture of the drug’s distribution in the body. In the experimental implant, small amounts of scandium radionuclides – 46Sc and 48Sc, were also detected, emitting sufficiently hard gamma radiation, which can pose a problem for patient dosage determination. The advantages of using titanium-47 with an enrichment of over 95 %, economically available, have been demonstrated, allowing for high radiochemical yields of 47Sc, sufficient for therapy. Conclusion: The 3D printing technology allows the production of a customized applicator for brachytherapy of specific dimensions and the delivery of arbitrarily-shaped sources to the tumor area for personalized therapy of oncological diseases. When implanting sources based on titanium alloys activated in a neutron flux of a research nuclear reactor, the radionuclide scandium-47 exhibits the highest activity.

On the Saturation of the Instability of Induced Ordinary Microwave Scattering in the Edge Transport Barrier of a Tokamak during Electron Cyclotron Resonance Heating of a Plasma

The saturation of the low-threshold parametric decay instability of an ordinary wave during electron cyclotron resonance heating in the edge transport barrier of a tokamak due to the stochastic damping of a daughter two-dimensionally localized oblique Langmuir wave has been considered. It has been found that the saturation of instability in modern devices occurs at a relatively low level and does not affect the energy balance during plasma heating. It has been shown that the efficiency of nonlinear pump under expected conditions of microwave power injection in the International Thermonuclear Experimental Reactor will exceed the maximum efficiency of stochastic damping, which will result in the breaking of amplitude-dependent saturation and can significantly modify the power deposition profile.

Best Estimate Plus Uncertainty Analysis of Station Blackout Transient in Pool-Type Research Reactor

Simulation of a station blackout (SBO) scenario was carried out for an open pool–type nuclear research reactor. The SBO transient was analyzed using the best estimate (BE) thermal-hydraulic code RELAP5/MOD3.2 to evaluate the performance of safety systems and inherent thermal inertia provided by the reactor pool in ensuring adequate core cooling during a prolonged SBO condition lasting up to 7 days. This encompasses assessment of cooling provided by battery-operated auxiliary pumps in the initial phase followed by setup of the natural convection cooling mode for the extended period. Best Estimate Plus Uncertainty (BEPU) methodology was applied for assessment of safety margins. This involved estimation of required simulations using the Wilks first-order formulation to achieve results within the tolerance limit of 95/95. Identification of relevant uncertainties and their propagation was carried out; subsequently, a case matrix for 59 simulation runs was generated using the Latin hypercube sampling method. The upper/lower bounds of uncertainty results were analyzed and compared with the BE code results. Later, sensitivity analysis was carried out using sensitivity coefficients generated using the Pearson and Spearman coefficient. The results show that the values of the crucial thermal-hydraulic parameters obtained with the tolerance limit of 95/95 met the acceptance criteria, with adequate safety margins.

Multiphysics and multiscale simulation for an experimental sodium-cooled fast reactor

This work aims to explore a new upscaling methodology to describe heat transfer simultaneously coupled with the neutronic process, in the core of a sodium-cooled reactor. The upscaling method, as demonstrated in this work, allows the development of models that describe reactor-scale phenomena, starting at the scale of a cell, composed of a nuclear fuel pin and liquid metal. In principle, this methodology allows high fidelity in the description of these multiphysics and multiscale phenomena. The main result is the two-upscaled energy no-equilibrium model to describe the temperature distribution of the fuel and coolant. These upscale models are described in terms of the upscaled heat transfer coefficient, which is presented and developed in this work, for the Experimental Sodium-cooled Fast Reactor. The use of the new proposed models/methodology leads good results within lower computational times compared with CFD codes, which is useful when computations have to be done a large number of times, for instance for sensitivity analyses to parameter choices in the conception phase of the reactor, and other applications such as plant analyzers, and for operator training in full-range simulators.

SELECTING OF MATRIX COMPOSITION FOR IMMOBILIZATION OF IRRADIATED FUEL OF THE IGR REACTOR

Impulse Graphite Reactor (IGR) is a unique nuclear installation namely research reactor with the core consisting of uranium-graphite fuel elements with 90% enrichment by U235. Possible immobilization of the first core taken from the reactor in 1967 is studied here as part of the IGR reactor conversion to low-enriched uranium fuel. Irradiated fuel is related to the category of mediumactive waste according to the content of beta-emitters. Expected value of highly-enriched uranium fuel (HEU) burnup rate was less than 1 % mass. The experts from the Institute of Atomic Energy (IAE) proposed the way how to immobilize irradiated IGR reactor fuel into the cement matrix. First, the fuel is suggested to dilute with depleted uranium in order to lower the enrichment up to the <20 mass. % by U235. The properties of the matrix should meet the engineering criteria ruled by international and national standards. So, according to aforesaid, The IAE team is facing an actual problem of selection the matrix and understanding its characteristics.The work was aimed at selection of composition of cement mortar for making robust matrix to immobilize irradiated IGR reactor fuel.The paper describes techniques and results of the time identification (start) of solidification, gradual volume change, water separation, velocity and uniform cement mortar.

Ensuring radiation safety during dismantling, transportation and long-term storage of the SM-3 research reactor core

Described shortly here is a procedure of demounting, removal, transport and long-term storage of the SM-3 core, based on the previous experience of reactor refurbishment undertaken in 1991. Prior to performing refurbishment, computations and calculated data analysis were performed to prove radiation safety of this work, which included estimation of the activity level for activation products in the structural materials of the nuclear research reactor core and the radiation conditions at different stages of its handling. As evidenced by the calculated data, the activity of the main dose-forming radionuclide 60Co attains equilibrium in about 12 years of radiation exposure. Taking into account the fact that the time period between two refurbishments was longer than 12 years, the calculated values of the equivalent dose rate were normalized to the radiation monitoring data obtained during the previous refurbishment, taking into account the calculated activity of 60Co radionuclide. The normalization made it possible to confirm reliability of estimates. The obtained activity data of activation products and taking into account the time spent during the SM-3 refurbishment in 1991, the radiation impact on personnel was estimated. Calculated values of the anticipated effective radiation exposure doses to the personnel engaged in the refurbishment revealed that the main limits of the personnel radiation exposure established in accordance with NRB-99/2009 were not exceeded. Comparison of the results of calculating the equivalent dose rate with the results of radiation monitoring at various points allowed us to establish that during the calculation and analytical justification of the radiation safety of work, the assessment of reflected radiation was significantly underestimated. But the radiation monitoring data, personal radiation monitoring, as well as recorded data of automatic radiation monitoring system show that all work was performed in compliance with the requirements of regulatory documents in the field of radiation safety.