Aqueous Homogeneous Reactor Research Articles

Thermal-hydraulic characterization of a 50-kW medicinal-aimed aqueous homogeneous reactor (MAHR)

This paper analyses the thermal–hydraulic performance of a proposed 50-kW aqueous homogeneous reactor (AHR) aimed at producing Mo-99. The reactor thermal–hydraulic characteristics are analyzed using ANSYS-FLUENT software and Relap code, and the results are validated through experiments conducted by a prototypic cylindrical sector of the reactor vessel. Boundary conditions in the CFD simulation are set up based on Relap modelling, while CFD outputs are validated by experimental results achieved using the laboratory model. Then, various parameters, including coolant mass flow rate, first cycle pipe diameter, coolant outlet temperature, average fuel solution temperature, and the elevation difference between the reactor and heat exchanger, are optimized using the least-squares optimization method. Results demonstrate that the heat removal system provides sufficient cooling capacity to ensure stable operation of the proposed AHR core by preventing fuel solution overheating and, consequently, boiling (void formation) in the active fuel solution.

Design and simulation of neutron radiography system for an aqueous homogeneous solution reactor

AHR reactors are known for the production of radiopharmaceuticals. The ARGUS reactor is one of the most famous of these reactors, which has been designed originally for radioisotopes production. The ARGUS reactor produces a suitable neutron flux that can be used for other applications such as neutron radiography (NR). Here, a NR system is designed step by step using the Monte Carlo method. The length of the collimator, divergence angle, diameter of the aperture, and thermal column dimension are designed. The designed aperture and thermal column are selected from boron carbide and heavy water, respectively. The depth of the thermal column is changed until the ratio of thermal neutrons to total neutrons (TNC) reached above 90%. The hole inside the thermal column is designed and internal lining is done. The inner lining is made of cadmium and had a thickness of 1 mm, and the space between the inner and outer lining is filled with lead. To reduce the gamma dose to acceptable values, the position of the collimator is changed instead of using a gamma filter. By changing the depth of the hole and creating the convergence angle to the hole, the flux of thermal neutrons will be 1.8E+6 n.cm−2s−1 that satisfy the IAEA suggested values for NR. Then, the uniformity of the radiation field of the neutron flux output from the collimator is obtained. At the end, images of SI and BPI standards are recorded. The good quality and contrast of these images showed the acceptable design of the NR system for the ARGUS reactor.

Research on the Formation Conditions and Preventive Measures of Uranium Precipitates during the Service Process of Medical Isotope Production Reactors.

This study focuses on the Medical Isotope Production Reactor (MIPR), an aqueous homogeneous reactor utilized for synthesizing medical isotopes like 99Mo. A pivotal aspect of MIPR's functionality involves the fuel solution's complex chemical interactions, particularly during reactor operation. These interactions result in the formation of precipitates, notably water filamentous uranium ore and columnar uranium ore, which can impact reactor performance. The research presented here delves into the reactions between liquid fuel uranyl nitrate and key radiolytic products, employing simulation calculations complemented by experimental validation. This approach facilitates the identification of uranium precipitate types and their formation conditions under operational reactor settings. Additionally, the article explores strategies to mitigate the formation of specific uranium precipitates, thereby contributing to the efficient and stable operation of MIPR.

Characterization of neutronic parameters and radiation shielding design for an aqueous homogeneous reactor

In this research, an aqueous homogeneous reactor is simulated and its neutronic parameters are investigated. Due to the negative reactivity caused by temperature change and void formation, the cold excess reactivity is considered equal to 4 βeff. Also, calculations of radiation shielding and neutronic characterization of the reactor are performed. For different heights of uranyl sulfate fuel, the effective multiplication factor (keff) of the reactor is calculated under cold zero power and hot full power conditions to assess the required excess reactivity. The amount of fuel burnup and changes in the keff are investigated for 500 days, finally 200 operating days are chosen as the working duration of the reactor. The control rods worth (CRW) for different diameters of boron carbid and reactor safety parameters including shot down margin (SDM) and safety reactivity factor (SRF), temperature and void coefficients of reactivity are analyzed. In order to design a suitable radiation shield for this reactor, two different shields are considered. The first shield is a combination of polyethylene thicknesses and barite concrete, and the second shield is made only of barite concrete. The total thickness of the first and second shield was 230 and 180 cm, respectively. Due to the problems of using polyethylene (e.g. problems related to machining, shield dimensions and low melting temperature), the second shield is chosen as the suitable radiation shield for this reactor. Using this radiation shield, the total dose rates of neutron and gamma rays satisfy the ICRP dose limits.

Comparative Analysis of Turbulence Models for Thermal-Hydraulic Simulations in Aqueous Homogeneous Reactors

This article presents a comparative study of various turbulence models applied in the context of thermal-hydraulic simulations for liquid fuel reactors, specifically Aqueous Homogeneous Reactors (AHR) using Computational Fluid Dynamics. The objective was to assess the suitability of the turbulence models by comparing their results with data obtained from Large Eddy Simulation (LES). For that purpose, was compared the flow behavior predicted using the k-ε, SST, GEKO, DES, SBES, and LES turbulence models. The calculations were carried out in a simplified computational model derived from a pre-existing three-dimensional AHR conceptual design. By utilizing this simplified model, the study aimed to focus on the computational differences between the turbulence models, while minimizing the influence of other factors. The calculation results revealed that the k-ε model exhibited significant discrepancies with the LES, with relative differences for the fuel solution maximum temperature reaching up to 75 %. Among the remaining RANS models, the Shear Stress Transport (SST) model demonstrated the best compromise between accuracy and computational efficiency, with differences below 5 % and requiring only 1/5th of the time, compared to the LES model. The Scale-Resolving Simulation (SRS) models,DES and SBES, provided a more comprehensive description of flow behavior and results closer to LES, albeit with higher computational demands. Between these two models, only the DES model exhibited relative differences below or equal to 1 % compared to the LES model for the studied thermohydraulic parameters.

Neutronic Evaluation of Using a Thorium Sulfate Solution in an Aqueous Homogeneous Reactor

Radioisotope 99 Mo is one of the most essential radioisotopes in nuclear medicine. Its production in an Aqueous Homogeneous Reactor (AHR) could be potentially advantageous compared to the traditional technology, based on target irradiation in a heterogeneous reactor. An AHR conceptual design using low-enriched uranium for the production of 99 Mo has been studied in depth. So far, the possibility of replacing uranium with a non-uranium fuel, specifically a mixture of 232 Th and 233 U, has not been evaluated in the conceptual design. Therefore, the studies conducted in this article aim to evaluate the neutronic behavior of the AHR conceptual design using thorium sulfate solution. Here, the 232 Th- 233 U composition to guarantee ten years of operation without refueling, conversion ratio, medical isotopes production levels, and reactor kinetic parameters were evaluated, using the computational code MCNP6. It was obtained that 14 % 233 U enrichment guarantees the reactor operation for ten years without refueling. The conversion ratio was calculated at 0.14. The calculated 99 Mo production in the AHR conceptual design resulted in 24.4 % higher with uranium fuel than with thorium fuel.

Implementation of a Multi-cell Approach in the Multi-Physics Calculations of an Aqueous Homogeneous Reactor

Nowadays, the Aqueous Homogeneous Reactor (AHR) technology is under study for its use in the production of medical isotopes. This technology has proven to be potentially advantageous for this purpose due to its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. Among the studies being carried out is computational modeling and simulation, which represents a key technology in the pursuit for improvements in efficiency, safety, and reliability of these systems. This paper aims to expand upon a previous AHR computational model through the implementation of a multi-cell approach for the improvement of the calculations using a methodology for the multi-physics and multi-scale coupling of the neutronic and thermal-hydraulic codes. It was found that these additions to the original model cause a small change to the overall reactor behavior. Thermal-hydraulic parameters such as, average fuel solution temperature and velocity, gas volume fraction and average radiolytic gas bubbles velocity undergo a variation of 0.161 °C, 0.0009 m/s, 0.015% and 0.0003 m/s. In contrast, significant local differences were obtained mainly for the fuel solution temperature and radiolytic gas bubbles volume fraction. It was verified that a simplified AHR computational model consisting of a 20° section of the fuel solution is able to adequately reproduce the results of the full AHR computational model.

Multi-physics evaluation of the steady-state operation of an Aqueous Homogeneous Reactor for producing Mo-99 for the Brazilian demand

The studies summarized in this paper aims to predict the steady state operation of a low-enriched uranium fuel ARGUS type aqueous homogeneous reactor for producing 99Mo to meet the domestic demand of Brazil through a coupled multi-physics (Neutronics + Thermal-hydraulics) evaluation. The coupled multi-physics evaluation included aspects related to the neutronic behavior such as fission induced energy deposition profile, medical isotopes production; and the thermal-hydraulic behavior such as temperature, velocities and gas volume fraction profiles. The methodology followed for the multi-physics and multi-scale coupling of the neutronic and thermal-hydraulic codes (MCNP + ANSYS-CFX), discussed in detail in this paper, represent one of the main outcomes of the current study. The methodology was tested for two different operating configurations of the ARGUS reactor, the original high-enriched uranium configuration used since 1981, and the new low-enriched uranium configuration after the conversion process during 2012-2014. The calculations carried out showed that the reactor, in the studied configuration, is able to produce 246.5 six days Curie of 99Mo in operation cycles of five days. Which is equivalent to more than a third of the estimated Brazilian demand for 2025.

New advances in the computational simulation of Aqueous Homogeneous Reactor for medical isotopes production

Aqueous Homogeneous Reactors (AHRs) or simply solution reactors present nowadays a promising alternative to produce medical isotopes, especially 99Mo. The AHR medical production concept has been proposed to produce medical isotopes directly in the fuel solution, resulting in a potentially competitive alternative in comparison with the solid target irradiation method in heterogeneous reactors. Furthermore, the utilization of AHRs for medical isotopes production has been strengthened because of the successful operation of the ARGUS reactor since 1981 and its conversion to low-enriched uranium (LEU) fuel during 2012-2014. Those successes positively influenced in the decisions to construct a Proof-Of-Concept production site based on the ARGUS operational experience in Sarov (500 km from Moscow) and to restore the Argus-FTI at the Umarov Physical and Technical Institute in Dushanbe, Tajikistan. However, demonstrating the viability of the AHRs for medical isotopes requires solving several significant challenges related with the safe operation of these reactors. Consequently, not only for the design, licensing and safe operation of the AHRs, but also for the prediction of accident scenarios it is very important to be able to simulate and predict the behavior of the fuel solutions through a group of relevant physical parameters. Accordingly, this paper aims to show the advances made to improve the predictive capabilities during the multi-physics computational modeling of AHRs.

A Method for Calculating the Number of Fissions in the VIR-2M Reactor from Measured Yields of Fission Products

A numerical method for calculating the fission rate in an aqueous homogeneous nuclear reactor from the yields of fission fragments is proposed. Fission fragment yields are measured by spectroscopically analyzing the fuel solution. Using this method, numbers of fissions in the core of the VIR-2M nuclear reactor are determined for experimental runs throughout its 17-year-long operation. The number of uranium fissions per 1 MJ of energy released in the VIR-2M core is estimated at 3.13 × 1016.

Conceptual design and uncertainty analysis of a typical 50 kW aqueous homogeneous reactor aimed for medical isotope production

Production of medical isotopes such as 99Mo, 90Sr, 131I and other fission products using Aqueous Homogeneous Reactors (AHRs) offers several advantages in comparison to other methods. In this method, after about one week operation, required isotopes can be directly extracted from the solution and the remaining fuel solution will be returned to the core for the next campaign. Moreover, AHRs are inherently safe and reliable due to their large negative reactivity feedbacks. This paper demonstrates technical features and conceptual scheme of the Iranian homogeneous reactor named RAHA, which is mainly used for the production of 99Mo/99mTc. Neutronic and thermal-hydraulic characteristics of the reactor were analyzed with MCNPX-2.6 and RELAP5/Mod3.3 codes, respectively. It was assumed that the reactor is fueled by Uranyl Nitrate, and is able to remove generated heat using natural circulation. The coolant circulation velocity was obtained as 6.5 cm/s. Four different types of neutron absorbers were considered as control rod material, and BORA resin with a diameter of 1.4 cm and Stainless-Steel cladding was chosen. As a new characteristic which may improve safety of the system, the possibility of reactor startup without any external neutron source for the case of fresh fueled core was assessed. Finally, effects of the possible uncertainties on some physical and material properties of the core were investigated and sensitivity of the effective multiplication factor relative to the uncertain parameters was also extracted using Spearman's formula.

Rod Ejection and Drop Accident Analysis of Aqueous Homogeneous Solution Reactor

Aqueous Homogeneous Solution Reactor concept has been proposed for producing medical isotopes(Medical Isotope Production Reactor-MIPR). However, there are several difficulties in transient calculation of aqueous homogeneous solution reactors. First, there are no assemblies in the core which is different from the traditional reactor core. Second, the operation of aqueous solution reactor at a power of 200kW will generate radiolytic-gas bubbles. The void volume created by these bubbles in the solution core will introduce a strong negative reactivity feedback. Third, the complex structure of the coolant pipes immersed in fuel solution requires unstructured neutron diffusion calculation methods. Therefore, analytic basis functions expansion nodal method for arbitrary triangular-z node is established to solve the complex structure geometry neutron diffusion equation. Based on this, a software named TABFEN-K has been developed to solve the three-dimensional space-time neutron kinetic equations. Then，TABFEN-K code is used for typical accident analysis of a solution reactor. A simplified geometry model, bubbles generation model , thermal conduction model and and cross section feedback model are given in this paper. A software called TABFEN-MIPR is developed and used for the simulations of the control rod ejection and drop. The same characteristics in the transient process with the results from literatures are obtained.

Neutronic evaluation of the steady-state operation of a 20 kWth Aqueous Homogeneous Reactor for Mo-99 production

Nowadays, 99mTc is the most common radioisotope used in nuclear medicine, with up to 30–40 million procedures worldwide every year. Furthermore, medical diagnostic imaging techniques using 99mTc represent approximately 80% of all nuclear medicine procedures. Currently, 99mTc is almost exclusively produced from the beta-decay of its 66-h parent 99Mo. The 99Mo production in an Aqueous Homogeneous Reactor (AHR) is potentially advantageous because of its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This paper studies an AHR conceptual design using low-enriched uranium for the production of 99Mo. Aspects related to the neutronic behavior such as critical height, medical isotopes production, uranium consumption, plutonium production and the reactivity feedback introduced in the solution by the volumetric expansion of the fuel solution were evaluated using the computational code MCNPX version 2.6e. In addition, important reactor kinetic parameters such as the effective delayed neutron fraction, βeff, and mean neutron generation time, Λ, were calculated. A benchmarking exercise was solved using available results of critical experiments performed at the Russian Research Center “Kurchatov Institute”. The neutronic calculations demonstrated that the reactor is able to produce 99 six-day Ci of 99Mo in operation cycles of five days. The reactivity feedbacks introduced by the volumetric expansion of the fuel solution is at about −1460 pcm, which represents approximately the 48% of the planned initial reactivity reserve in the core. The calculated effective delayed neutron fraction and the average mean neutron generation time were 785 pcm and 134.08 µs, respectively.

Effects of some calculation parameters on the computational modelling of temperature, velocity and gas volume fraction during steady-state operation of an aqueous homogeneous reactor

This paper is part of the ongoing efforts to contribute to the thermal-hydraulic analysis of one of the most promising alternatives to produce medical isotopes and meeting current and future demand for 99Mo: the use of Aqueous Homogeneous Reactors (AHRs). In this paper, the effects of some calculation parameters like mesh refinement, time step size, turbulence models, transient schemes and numerical advection scheme on the computational modelling of key parameters of an AHR steady-state operation have been investigated. For this purpose, a 75 kWth AHR conceptual design based on the ARGUS reactor, six meshes, five time step sizes, three different models for solving flow problems, three numerical advection schemes and the available transient schemes were used in the simulations. The numerical simulations were carried out using the Computational Fluid Dynamic (CFD) code ANSYS CFX 14. The results of the CFD simulations allow developing a detailed and improved CFD model of the AHR core on which the effects of the investigated calculation parameters are quantifiable.

Effects of some calculation parameters on the computational modelling of temperature, velocity and gas volume fraction during steady-state operation of an aqueous homogeneous reactor

This paper is part of the ongoing efforts to contribute to the thermal-hydraulic analysis of one of the most promising alternatives to produce medical isotopes and meeting current and future demand for 99Mo: the use of Aqueous Homogeneous Reactors (AHRs). In this paper, the effects of some calculation parameters like mesh refinement, time step size, turbulence models, transient schemes and numerical advection scheme on the computational modelling of key parameters of an AHR steady-state operation have been investigated. For this purpose, a 75 kWth AHR conceptual design based on the ARGUS reactor, six meshes, five time step sizes, three different models for solving flow problems, three numerical advection schemes and the available transient schemes were used in the simulations. The numerical simulations were carried out using the Computational Fluid Dynamic (CFD) code ANSYS CFX 14. The results of the CFD simulations allow developing a detailed and improved CFD model of the AHR core on which the effects of the investigated calculation parameters are quantifiable.

Neutronic and thermo hydraulic analysis of a modeled subcritical uranyl nitrate aqueous reactor driven by 30-MeV protons

Recently aqueous homogeneous reactors have been taken in the spotlight again because of their good potential for radioisotope product since ARGUS exhibits such successful experience. Accelerator-driven types of such aqueous homogeneous reactors are being investigated and designed due to some their prominent advantages than the used ones by operation in critical level. The present work discusses neutronic and thermo-hydraulic performance of a modeled accelerator-driven aqueous homogeneous reactor. 30MeV protons with 150μA current were considered to induce fission process inside the core involving uranyl nitrate solution. Monte Carlo-based computational code was used to model such subcritical system. FLUENT code was used to determine the temperature distribution inside the operated subcritical core with an offered simplified geometry. The carried out calculations showed that the modeled system with external beam window positioning experiences 4.683kW fission-related power. A neutron flux of the order of 1011n/s.cm2 is available inside the subcritical core. Enough negative reactivity coefficient of the modeled core moreover its subcritical level operation assures the core inherent safety. The system can be used to produce an adequate yield of different radioisotopes per week. Thermo-hydraulic system used for the modeled core could efficiently remove the aqueous core produced heat and keep the core temperature below 70°C.

Thermal-Hydraulics Study of a 75 kWth Aqueous Homogeneous Reactor for 99Mo Production

Tc99m is a very useful radioisotope, which is used in nearly 80% of all nuclear medicine procedures. Tc99m is produced from 99Mo decay. A potentially advantageous alternative to meeting current and future demand for 99Mo is the use of Aqueous Homogeneous Reactors (AHR). In this paper, a thermal-hydraulics study of the core of a 75 kWth AHR conceptual design based on the ARGUS reactor for 99Mo production is presented. As the ARGUS heat removal systems were designed for working at 20 kWth, the main objective of the thermal-hydraulics study was evaluating the heat removal systems in order to show that sufficient cooling capacity exists to prevent fuel solution overheating. The numerical simulations of an AHR model were carried out using the Computational Fluid Dynamic (CFD) code ANSYS CFX 14. Evaluation shows that the ARGUS heat removal systems working at 75 kWth are not able to provide sufficient cooling capacity to prevent fuel solution overheating. To solve this problem, the number of coiled cooling pipes inside the core was increased from one to five. The results of the CFD simulations with this modification in the design show that acceptable temperature distributions can be obtained.

An extension of the point kinetics model of MIPR to include the effects of pressure and a varying surface height

This work aims to expand upon a previous point kinetics model of an aqueous homogeneous reactor and make several additions to it including the effects of average pressure, plenum gas composition and temperature, vertical advection of solution due to voidage and thermal expansion and a neutron flux profile which varies in both time and height from the reactor base. It is found that these additions cause a small change to the reactor behaviour but the main advantage is to allow the modelling of different scenarios including failure of the gas management system which normally removes hydrogen and fission product gases from the plenum gas. Such a failure is seen to lead to a rapid increase in hydrogen content with implications for reactor safety. The model is also used to simulate the CRAC-43 experiment and, with the addition of simple models for the delay of radiolytic gas production due to the need to saturate the fuel solution and sloshing of the fuel solution, is found to approximate the experimental results well.

Analysis of <sup>99</sup>Mo Production Capacity in Uranyl Nitrate Aqueous Homogeneous Reactor using ORIGEN and MCNP

99m Tc is a very useful radioisotope in medical diagnostic procedure. 99m Tc is produced from 99 Mo decay. Currently, most of 99 Mo is produced by irradiating 235 U in the nuclear reactor. 99 Mo mostly results from the fission reaction of 235 U targets with a fission yield about 6.1%. A small additional amount is created from 98 Mo neutron activation. Actually 99 Mo is also created in the reactor fuel, but usually we do not extract it. The fuel will become spent fuel which is a highly radioactive waste. 99 Mo production system in the aqueous homogeneous reactor offers a better method, because all of the 99 Mo can be extracted from the fuel solution. Fresh reactor fuel solution consists of uranyl nitrate dissolved in water. There is no separation of target and fuel in an aqueous homogeneous reactor where target and fuel become one liquid solution, and there is no spent fuel generated from this reactor. Simulation of the extraction process is performed while reactor in operation (without reactor shutdown). With an extraction flow rate of 3.6 L/h, after 43 hours of reactor operation the production of 99 Mo is relatively constant at about 98.6 curie/hour. /hour. Received: 11 January 2014; Revised: 18 February 2014; Accepted: 28 February 2014

The application of polynomial chaos methods to a point kinetics model of MIPR: An Aqueous Homogeneous Reactor

This paper models a conceptual Medical Isotope Production Reactor (MIPR) using a point kinetics model which is used to explore power excursions in the event of a reactivity insertion. The effect of uncertainty of key parameters is modelled using intrusive polynomial chaos. It is found that the system is stable against reactivity insertions and power excursions are all bounded and tend towards a new equilibrium state due to the negative feedbacks inherent in Aqueous Homogeneous Reactors (AHRs). The Polynomial Chaos Expansion (PCE) method is found to be much more computationally efficient than that of Monte Carlo simulation in this application.