Speciation and estimation of radioiodine in reactor coolant water

The species of radioiodine in the primary coolant water of the heavy water moderated, heavy water cooled 100 MW research reactor have been identified. It was observed that IO ₃ ^? was the major species in the reactor coolant during reactor operation and I^? was the major species during shutdown. Organic and elemental forms amount only to less than 2% of total radioiodine. A simple method was developed for the estimation of gross iodine activity in reactor coolant water. The method involves the separation of all inorganic forms of iodine into a photographic film consisting of a thin layer of silver halide. The iodine in the film was estimated by gross counting of the film in a Geiger-Müller counter. Gross iodine activity in the reactor coolant samples estimated by the present method were in agreement with that obtained by direct γ-spectrometry with a Ge detector. It is concluded that the method can be used for the routine estimation of radioiodine in reactor coolant water.¹³³I/¹³¹I and¹³⁵I/¹³¹I ratios in the film were estimated and found to be useful in identifying split rod conditions in the reactor.     

Determination of in-core power in low energy research reactors by measurement of16N and18F in the primary coolant

The measurement of¹⁶N and¹⁸F activity in the primary coolant of the JASON Argonaut reactor has been used to monitor in-core reactor power. The¹⁶N is produced by the¹⁶O(n, p)¹⁶N reaction and the 6.1 MeV photopeak was measured on-line using a BGO detector adjacent to the primary coolant circuit. These data provided a relative measure of power stability during steady state operation and a measure of linearity at different power levels. The¹⁸F is produced in the primary coolant by the¹⁸O(p, n)¹⁸F reaction and aliquots of primary coolant were sampled from the reactor dump tank for off-line radiochemical analysis. The¹⁸F was separated as trimethylfluorosilane and the activity was determined by measurement of the 0.511 MeV annihilation photopeak using a NaI(TI) detector. The measured¹⁸F activity was used to determine actual in-core reactor power using both ab-initio calculations and by comparison of results with a calibrated power reactor. The¹⁸F data also provided a method of nomalising the¹⁶N data for direct monitoring of in-core reactor power in JASON.     

Application of BGO detector to monitor in situ16N concentration in a reactor pool

A high efficiency BGO detecting system coupled to a coolant pumping device was used to monitor on-line the 6.13 MeV high energy gamma-ray, which is emitted from the 7.1 s half-life of¹⁶N activated by¹⁶O(n, p) reaction in a nuclear reactor. The system has been demonstrated effectively to monitor the real-time reactor power level as well as the space-time neutron flux distribution in reactor coolant.     

Recovery of plutonium from aqueous waste solutions using porous zirconia spherical particles

Based on the Linssi database and UniSampo/Shaman software, an automated analysis platform has been setup for the analysis of large amounts of gamma-spectra from the primary coolant monitoring systems of a CANDU reactor. Thus, a database inventory of gaseous and volatile fission products in the primary coolant of a CANDU reactor has been established. This database is comprised of 15,000 spectra of radioisotope analysis records. Records from the database inventory were retrieved by a specifically designed data-mining module and subjected to further analysis. Results from the analysis were subsequently used to identify the reactor coolant half-life of ¹³⁵Xe and ¹³³Xe, as well as the correlations of ¹³⁵Xe and ⁸⁸Kr activities.     

Plurality of radiation-chemical steady states during the radiolysis of water coolant in nuclear reactor

A computational study has been performed on the radiolysis of an aqueous solution that has the chemical composition and irradiation parameters typical of the primary coolant of pressurized water reactors. In this work, in contrast to previous studies on this subject, it has been taken into account that temporary fluctuations of the reactor radiation parameters, such as the composition of reactor n, γ]-radiation and the absorbed dose rate in the coolant, may occur in an actual real reactor during irradiation. The feasibility of spontaneous switching of the radiation-chemical system from one steady state with low concentrations of oxidizing products of radiolysis to another state with high concentrations resulting from a spike of the reactor radiation dose rate against the background of constant dose rate before and after the spike has been shown. The event triggering the switch is an increase in the concentration of oxygen and HO₂ (O₂^–) radicals during the spike, after which the positive feedback does not allow the system to return back to the initial steady state. In practice, this effect can be caused by phenomena associated with local irregularities in the reactor core: the composition and dose rate of reactor radiation, the presence of steam–gas bubbles due to local subcooled nucleate boiling of the coolant in the core, and the perturbation of the water radiolysis mechanism at the surface of fuel rods because of an increased contribution of heterogeneous reactions etc.     

Analysis of actinides in the primary coolant of a WWER-440 type reactor

A simple and effective method has been developed for the determination of the transuranium isotopes in primary coolant samples of WWER-440 type reactors. Membrane filters containing undissolved particles from about one liter coolant were destroyed by mineral acids, then americium-curium isotopes and neptunium-plutonium isotopes were coprecipitated sequentially with NdF₃ based on their different redox behavior. Alpha-spectra were determined by silicon semiconductor detectors connected to multichannel analyzers. Activity ratios measured in a reactor unit with a failed fuel element were compared with calculated values of the same type of reactor under similar conditions. Parameters of the defective fuel (burnup, original enrichment) could be estimated.     

Determination of radioactive cobalt in reactor coolant water by solvent extraction

A rapid extraction separation method is described for the radiochenaical determination of cobalt-58 and cobalt-60 in reactor coolant water. After adjustment of the pH of the sample water to 5.0-5.5, cobalt diethyldithiocarbamate is extracted with benzene. Other nuclides, e.g., manganese-54 and -56, copper-64 and iron-59, which are usually present in reactor coolant water, are also extracted together with cobalt. However, they can be readily removed by washing the extract with mercury(II) chloride solution. The cobalt-58 and -60 activities are measured by gamma counting of an aliquot of the washed extract. Trace amounts of radioactive cobalt in 500 ml of sample water can be quantitatively extracted without the use of carrier. The separation could be finished in 15 min and an average recovery of 99.5 % was obtained, with a relative standard deviation of 1.4% (25 experiments).     

Xenon isotopic signature study of the primary coolant of CANDU nuclear power plant to enhance CTBT verification

To support interpretation of observed atmospheric ¹³⁵Xe, ¹³³Xe, ^133mXe and ^131mXe, a database of xenon radioisotope in the primary coolant of CANDU reactors has been established. This database is comprised of 40000 records of high-quality xenon radioisotope analyses. Records from the database were retrieved by a specifically designed data-mining module and subjected to further analysis. Results from the analysis were subsequently used to study isotopic ratios of observed xenon radioisotopes in the CANDU reactor primary coolant. These studies provided novel and practical information on the characterization of CANDU reactor xenon radioisotope releases, which can be used to discriminate between reactor effluence and underground nuclear test releases.     

Krypton isotopic signature study of the primary coolant of CANDU nuclear power plant

To support interpretation of observed atmospheric krypton radioisotopes, a database of krypton radioisotope in the primary coolant of CANDU reactors has been established. This database is comprised of 40,000 records of high-quality ⁸⁹Kr, ⁸⁷Kr, ⁸⁸Kr and ^85mKr analyses. Records from the database were retrieved by a specifically designed data-mining module and subjected to further analysis. Results from the analysis were subsequently used to study isotopic ratios of observed krypton radioisotopes in the CANDU reactor primary coolant. These studies provided practical information on the characterization of CANDU reactor krypton radioisotope, which can potentially be used to discriminate between reactor effluent and fuel reprocessing for nuclear safeguard ⁸⁵Kr monitoring applications (Kalinowski et al., J Environ Radioact 73:203, 2004). The study also has some potential application to Fissile material cut-off treaty.     

Hot particles analysis originating from failed and damaged fuels

The increase of activities of fission products and transmutation products in the primary coolant of a nuclear power plant indicates the presence of fuel rod failures. The measurement of the activity concentration of the primary coolant was able to detect fuel failures in the reactor core. Microanalytical methods for examining individual hot particles have been developed and applied to fuel failure detection under normal operation conditions as well as during the severe fuel damage that occurred in the cleaning tank incident at Unit 2 of NPP Paks in April 2003. Several faulty fuel rods can be detected simultaneously by the characterization of individual hot particles originating from the primary water. The analysis of particles originating from the damaged fuels provides information relating to the dissolution process of the fuel debris.     

Determination of soluble chromium in simulated PWR coolant by differential-pulse adsorptive stripping voltammetry

An analytical method has been developed for the determination of dissolved chromium at concentrations less than 2 mug/l. in PWR coolant by differential-pulse adsorptive stripping voltammetry at a hanging mercury drop electrode. Concentrations above 2 mug/l. can be determined by appropriate dilution of the sample. The method is based on measurement of the current associated with reduction of a chromium(III)-DTPA (diethylenetriaminepenta-acetic acid) complex adsorbed at the surface of the mercury drop. The effects of boric acid, pH, DTPA concentration, accumulation potential and time were investigated together with the oxidation state of the chromium. No interference was observed from other transition metal ions expected to be present in PWR coolant. No alternative chemical technique of similar sensitivity was available for comparison with the results obtained in solutions containing <1 mug/l. chromium. Recoveries from simulated coolant solutions were greater than 95% and the relative standard deviations for single determinations were in the range 12-25%. The statistical limit of detection at the 95% confidence level was 0.023 mug/l. This method of analysis should prove valuable in corrosion studies and is uniquely capable of following the changes in soluble chromium concentration in PWR coolant that follow operational changes in the reactor.     

Electronic Simulation of Neutronic Transient Studies for PWR-900 MW(e)

To study different functionalities of the PWR-900 MW(e) reactor core, an electronic model of Simulation Program with Integrated Circuit Enfesied (PSPICE) for electronic circuit design was adopted to model the mathematical models. A classical breadboard electronically simulates the physical terminal characteristics of each network. The electronic simulator, parity simulator, for reactor core can simulate six-delayed neutron groups, single prompt neutron, neutron source, fuel channel, the coolant, reactor poisons, temperature feedback, control rods, safety rods, and boron injection. Positive/negative step or sin reactivity functions are two case studies were analyzed. The obtained results were found to be in a good agreement with the available data of El-Dabaa PWR 900 MW(e) basic simulator.     

Robust model-based predictive control of exothermic chemical reactor

A case study of the robust model-based predictive control (MPC) of an exothermic continuous stirred tank reactor (CSTR) with uncertain parameters is presented. Three robust MPC approaches are considered and the simulation results are compared in terms of quality of control performance and total consumption of coolant. The results reveal the main benefits of the considered approaches and confirm that the robust MPC can bring about a reduction in consumption of the coolant.     

Measurement of <Superscript>90</Superscript>Sr in primary coolant of pressurized water reactor

The present procedure for analysing ⁹⁰Sr combines the use of a non radioactive carrier with high pressure ion chromatography separation, conductivity detection of the carrier and optimized external counting by liquid scintillation. This improvement with respect to traditional methods led to a more rapid and efficient purification stage. The present work proves that activities of ⁹⁰Sr as low as 3 Bq/L can be measured in highly contaminated pressurized water reactor primary coolant matrix without any observed radiochemical interference. The approach shows promise for the analysis of other emitters of low energy radiation, or isotopes subject to high background or matrix effects in a PWR primary coolant.     

Migration of fission product137Cs in mud near a nuclear reactor

Due to the mismanagement of nuclear waste as well as heat exchanger degradation for the primary coolant of the one megawatt nuclear research reactor, the fission product¹³⁷Cs has been leaking to the environment ever since 1969. In the past thirty years, the long-lived¹³⁷Cs was accumulated and eventually trapped in the mud of the discharge pond right in front of the waste storage and the reactor facility. The distribution of¹³⁷Cs in mud was measured and contour-mapped to reveal the migration of trace levels of¹³⁷Cs in a period of three decades.     

Extraction and DFT study on the complexation of the methylammonium cation with a hexaarylbenzene-based receptor

In methods for quantification of ⁶³Ni, in e.g. reactor coolant water, a chemical separation is required due to ⁶³Ni being a pure beta emitter with limited means of quantification. ⁶⁰Co, a common radionuclide in reactor coolant water, is not completely separated with the commonly used separation procedure, and it is not resolved from ⁶³Ni in the beta spectrum. The separation method discussed in this work consists of TRU resin (Eichrom) and Ni resin (Eichrom). After running the separation procedure, depending on the initial activity of ⁶⁰Co, there may still remain enough ⁶⁰Co to interfere in the measurement of ⁶³Ni. The ⁶⁰Co interference is corrected for via a gamma spectrometric measurement. This correction may, depending on the ⁶³Ni/⁶⁰Co ratio, introduce a large contribution to the measurement uncertainty. The aim of this work was to evaluate the possibility to reduce the measurement uncertainty of ⁶³Ni measurements by adding a second Ni separation to the method. Double Ni separations were performed on reactor coolant water having a ⁶⁰Co activity much higher than the ⁶³Ni activity (⁶³Ni/⁶⁰Co = 0.01), in order to decrease the radioactivity of ⁶⁰Co in the sample. The measurement uncertainty of the ⁶³Ni measurement result was reduced by a factor of about three.     

Simultaneous radiochemical separation of fractions of radioisotopes of primary nuclear reactor coolant water

Discs coated with silver iodide, lanthanum oxalate and calcium-barium sulfate were simultaneously rotated in nuclear reactor coolant water samples for radiochemical separation of iodine isotopes, the group of radioactive rare earth elements and the group of radioactive alkaline earth elements. The reproducibility of the method is discussed.     

Application of a Direct Cooling Method in the Cationic Polymerization of Isobutylene

Inspired by the Leidenfrost phenomenon, a direct-cooled, general-purpose, low-temperature reactor has been developed and its applicability for polymerization was demonstrated. The Leidenfrost reactor is efficiently cooled by introducing relatively inexpensive liquid N₂, (LN₂), directly into the bulk of polymerization charges; the rapid evaporation of the coolant chills the system. The gaseous N₂, generated provides an inert gas blanket that protects moisture- and/or air sensitive charges. The installation of expensive cooling jackets or coils is superfluous and stirring is optional. The virtually complete recovery of cooling energy is possible. In the Leidenfrost reactor, the LN₂, heat sink is virtually in contact with the heat source so that the path of heat transfer is practically nil.     

Determination of radioactivity concentration in the primary coolant channel of a Pressurized Heavy Water Reactor

CdZnTe detector was used to monitor the radioactivity at several locations of primary heat transport (PHT) system of the Tarapur Atomic Power Station-III & IV reactor during shut down as well as operating condition of the reactor. Dose measurements at the same locations were also carried out using a teletector. MCNP code was used to simulate the detector efficiency for the required counting geometry. The efficiencies were then used to convert the experimental count rates (cps) to corresponding activity concentration (Bq/mL) of different radionuclides in coolant water. The total dose rates in these locations were also calculated by summing up the simulated dose rates of individual γ ray energy. The simulated and the measured total dose rates were found to be in reasonably good agreement.     

Rough localization of defective PWR fuel rods on the basis of the concentration ratio134Cs/137Cs in the primary coolant

The search for fuel assemblies with defective fuel rods by sipping tests involves expensive prolongations of the refuelling periods in nuclear power plants. It is attempted to reduce the number of fuel assemblies to be checked by sipping during the refuelling periods by means of rough localization of the defects on the basis of the fission product concentration ratio¹³⁴Cs/¹³⁷Cs in the primary coolant already during reactor operation. First results obtained in two examples of application at VVER type reactors are encouraging. The burnup of the defective assemblies could be correctly predicted from the cesium ratio in these two cases, which was confirmed by sipping tests after reactor shutdown.