Aqueous Homogeneous Reactor Research Articles

Simulated transient dynamics and heat transfer characteristics of the water boiler nuclear reactor – SUPO – with cooling coil heat extraction

The term “water boiler” reactor refers to a type of aqueous homogeneous reactor (AHR) that was designed, built and operated by Los Alamos in the 1940s. This was the first type of liquid fuelled reactor and the first to be fuelled with enriched Uranium. For security reasons the term “water boiler” was adopted and three versions were built: LOPO (for low power), HYPO (for high power) and SUPO (for super power) which were spherical shaped reactor vessels. The name was appropriate as the reactors appeared to boil although this was actually due to the release of radiolytic gas bubbles; although SUPO was operated during some studies close to the boiling point of uranyl nitrate. The final water boiler “SUPO” was operated almost daily as a neutron source from 1951 until its deactivation in 1974-23years of safe, reliable operation. Many of the key neutron measurements needed in the design of the early atomic weapons were made using LOPO, HYPO and SUPO. More recently SUPO has been considered as a benchmark for quasi-steady-state operation of AHRs with internal cooling structures.This paper presents modelling and analysis of the coupled neutronic and fluid time dependent characteristics of the SUPO reactor. In particular the quasi-steady-state dynamics of SUPO have been investigated together with its heat transfer characteristics. In the simulations presented the SUPO reactor is modelled using the spatially dependent neutron/multiphase CFD simulation tool, FETCH, at a quasi-steady-state power of 25kW. SUPO also possessed a cooling coil system that fed cooling water through the reactor for the extraction of the fission and decay heat. This cooling system, and the heat extraction, is modelled in the simulations using a new sub-modelling approach that is detailed here. The results from this simulation, such as gas fraction, gas generation rate, coolant rate and average temperature, are compared against the available experimental information.

Calculation of the Void Fraction and Void Coefficient in an Aqueous Homogeneous Reactor

The aqueous homogeneous reactor has assumed some importance in recent years as a potential medical isotopes production system. The kinetic behavior of such systems depends on the rate of generation of the radiolytic gas bubbles and the associated reactivity void coefficient. In this work we describe a method based on perturbation theory, and a simple description of bubble production, for deriving a value of the void coefficient of reactivity. It is shown that, in the small void fraction limit, the void coefficient is dependent only on the system properties and does not depend on power level or the bubble properties. Values are given for the void coefficient for a range of parameters assuming that the voids are distributed in three different ways, i.e., uniformly, proportional to power, and equal to the distribution due to buoyancy. The results are in reasonable agreement with values obtained by others, using more detailed methods, based on the reactors SILENE and MIPR.

FMSR: A code system for in-core fuel management calculation of aqueous homogeneous solution reactor

Aqueous homogeneous solution reactor is a promising concept for the production of medical isotopes. But some characteristics of aqueous solution reactors, such as no traditional assembly in the core, the gas bubbles’ generation in fuel solution, isotopes distillation, unstructured geometry, strong anisotropic scattering, etc., make the fuel management calculation very complicated. This study establishes a suitable calculation model for aqueous homogeneous solution reactors and developed an in-core fuel management calculation code FMSR (Fuel Management for Solution Reactors) based on the 3D transport solver DNTR. Numerical results indicate that FMSR can be used for the fuel management calculation of homogeneous aqueous solution reactor as a trial.

An interleaved approach to production of 99Mo and 89Sr medical radioisotopes

More than half of the world's annual production of radionuclides is used for medical purposes such as diagnostic imaging of diseases and patient therapy. Using aqueous homogeneous solution reactor technology, production quantities of medical radioisotopes 99Mo and89Sr, can be extracted from one reactor cycle. 99Mo may be produced directly from UO2SO4 uranyl sulfate in an aqueous homogeneous solution nuclear reactor in a manner that produces high purity radionuclides, making efficient use of the reactor's uranium fuel solution. The process is relatively simple, economical, and waste free, eliminating uranium targets. The short-lived radioisotope 99mTc is eluted from 99Mo for diagnostic imaging. Radioisotope 89Sr infusion is a therapeutic modality that reduces reliance on narcotic analgesia through palliation of metastatic bone pain caused by metastases of the cancer to the bone. Painful disseminated osseous metastases are common with carcinomas of the lung, prostate, and breast. Synergistic interleaving of two manufacturing processes, one producing 99Mo and another producing 89Sr in the same production cycle of an aqueous homogeneous solution reactor makes full and efficient use of the time for both the neutron irradiation stage and the extraction stage of each radionuclide. Interleaving the capture of 89Sr radioisotope with production processing of 99Mo radioisotope is achieved, since the extraction and subsequent elimination of radionuclide impurities occurs during separate parts of the reactor cycle. The process applies to either HEU or LEU nuclear fuels in an aqueous homogeneous solution reactor.

Dynamics of aqueous homogeneous pulse reactors

Dynamics of aqueous homogeneous pulse reactors

Spectrophotometry of Aqueous Homogeneous Reactor Fuel Solutions from 4 to 280°C and 0 to 930 psi(abs)

Aqueous homogeneous reactor fuel solutions have been examined spectrophotometrically in the wavelength range 0.3 to 1.2µ as a function of time, temperature, and overpressures of hydrogen and/or oxy...

Stress Corrosion Cracking of Austenitic Stainless Steel In Uranyl Sulfate Solutions

Abstract The effects of a number of variables on the stress corrosion cracking behavior of two different heats of Type 347 stainless steel were examined in simulated aqueous homogeneous reactor fuel solution containing low concentrations of chloride ions. The fuel solution consisted of uranyl sulfate, sulfuric acid and copper sulfate. Tests were conducted in the temperature range of 50 C to 101 C Although the two heats of stainless steel were nearly identical in composition, specimens from one heat were consistently more susceptible to cracking than specimens from the second heat. Chloride concentrations of 25 to 500 ppm were equally effective in producing cracks. The higher the dissolved oxygen content in solution, the more numerous and intensive were the cracks. Cracking was still observed, however, in the absence of oxygen. Uranyl sulfate concentration was an important variable; at concentrations of 0.40M and greater, no cracking took place with chloride present. Sodium dichromate was found to be an ef...

半均質炉と水性均質炉の開発研究

Studies on a high temperature gas cooled reactor with semi-homogeneous fuel were started in JAERI in 1958, when this reactor concept was taken up.Liquid bismuth as well as helium were considered as coolant. The main items of study under the JAERI program are described. An aqueous homogeneous reactor program in past several years is also reviewed. The main activities were critical experiments and analyses of a homogeneous heavy-water system, studies on behavior of a thorium oxide slurry and preliminary mock-up experiment for a boiling slurry reactor concept.

Analysis of the neutron flux fluctuations in an aqueous homogeneous subcritical reactor with circulating UO 2 particles

The neutron flux noise in the KEMA subcritical assembly, described by Kersten (1962), has been analysed by autocorrelation techniques. When a stable suspension was used, the oscillations showed smaller amplitude than with the unstable suspension. With the latter, characteristic periods of 8 and 16 sec were found. With the stable suspension only the 16 sec period was dominant.

Experiments with a subcritical reactor, in which an aqueous suspension of UO 2 particles is circulating

Measurements on source multiplication and on neutron flux variations (reactor noise) have been used to investigate rheological and reactor physical properties of an aqueous homogeneous subcritical reactor, in which a suspension of UO 2 particles is circulating at room temperature. It has been shown that both the colloid chemical properties of the suspension and the hydrodynamical conditions inside the core vessel have a strong influence on the behaviour of the reactor at steady operation. Interesting results have also been obtained under abnormal operating conditions. The effective delayed neutron fraction has been measured as a function of the circulating velocity and the result appeared to be in agreement with theoretical considerations.

Aqueous systems at high temperature—III investigations on the system UO 3CuONiOSO 3H 2O at 300°C

Solubility relationships at 300°C in the five-component system UO 3CuONiOSO 3H 2O, and its included aqueous four- and three-component systems, have been investigated. Also reported are a few data on the analogous D 2O system. In order to gain over-all information in a relatively short period of time, the experimental program consisted in determining ternary L + S, quaternary L + 2 S, and quinary L + 3 S curves at constant temperature for which, by definition, one, two and three solid phases, respectively, were in equilibrium with a liquid phase. For any set of experiments containing the same solid phases the SO 3 concentration was varied in order to give a solubility curve as a function of this variable. Accordingly, skeletal solubility data, obtained for the five-component system and its accessory four- and three-component systems, gave an over-all description of the solubility equilibria at 300° in the condensed system. Comments are made regarding the nature of the species in solution, the general shape of some of the solubility curves, the correlation of solid-liquid with liquid-liquid equilibria, and a specific application to aqueous homogeneous reactor fuels.

Aqueous systems at high temperature—II: Liquid-liquid immiscibility in the system UO 3SO 3N 2O 5H 2O above 300°C

Apparatus suitable for the observation of liquid-liquid immiscibility in solutions sealed in silica tubes is described. Immiscibility temperatures are reported for 0·04, 0·08 and 0·16 molal uranyl solutions in the system UO 3SO 3N 2O 5H 2O. A comparison of the effect of nitric and sulphuric acids on the immiscibility temperature of uranyl sulphate solutions on the basis of total added hydrogen ion shows that at low acidity nitric acid produces a greater increase in immiscibility temperature than sulphuric acid and at high acidity sulphuric acid becomes more effective than nitric acid. In solutions containing low concentrations of excess acid, the substitution of nitrate for sulphate raises the immiscibility temperature, but at high acidities the initial substitution lowers the immiscibility temperature. The data seem to indicate that there is an interdependence between acidity and complexing in this system. The data also show that partial substitution of nitrate for sulphate in a sulphate-based aqueous homogeneous reactor fuel at low added acidity may have some advantage in elevating the liquid-liquid immiscibility temperature. Finally, a significant effect of pressure in elevating the immiscibility temperature is reported.

Some Aspects of Fluid Fuel Reactor Development

Fluid fuel reactor development is reviewed and the advantages and disadvantages of this type of reactor are given. The technology of the Aqueous Homogeneous Reactor, the Aqueous Homogeneous Reactor Experiments, Molten Salt Reactors, and the Aircraft Reactor Experiment are discussed. (M.C.G.)

二領域均質炉の燃料サイクル

The ways of converting atwo-region aqueous homogeneous reactor fueled with 20% 235U enriched uranium into a 233U-breeder are discussed. Numerical solutions are obtained for the burn-up equations combined with a one-group criticality condition and constant total power out-put. “Batch method” and “Continuous method” are compared. In each case, time-dependent changes in the concentrations of individual isotopes, the number of uranium isotopes supplied to the core and removed from the blanket, chemical processing rate and the ratio of the power produced in the blanket to that in the core, as well as the effect of 233U concentration in the blanket on these quantities are discussed. We define “breeding efficiency” in this paper, which is believed to be a more accurate measure of whether a reactor is a breeder or not than the conventional breeding ratio.

Molten Fluorides as Power Reactor Fuels

A high-temperature reactor using molten uranium salts (Aircraft Reactor Experiment) was operated for a short time at the Oak Ridge National Laboratory. The reactor was of the circulating-fuel type, with a BeO moderator. The maximum outlet temperature achieved was greater than 1500°F (1100 K). It is believed that with further development the ARE could be a prototype for an economical uranium burner. Two very different schools of reactor design have emerged since the first reactors were built. One approach, exemplified by solid fuel reactors, holds that a reactor is basically a mechanical plant; the ultimate rationalization is to be sought in simplifying the heat transfer machinery. The other approach, exemplified by liquid fuel reactors, holds that a reactor is basically a chemical plant; the ultimate rationalization is to be sought in simplifying the handling and reprocessing of fuel. At the Oak Ridge National Laboratory we have chosen to explore the second approach to reactor development. The ORNL aqueous homogeneous reactor is the best-known embodiment of the liquid reactor approach; in a sense it represents the natural culmination of the aqueous reactor systems.