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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

Measurements of actinide nuclides in water samples from the primary circuit of a nuclear power plant

Weekly, consecutive primary coolant samples from a boiling water reactor have been analyzed for^{239, 240}Pu,²³⁸Pu+²⁴¹Am,²⁴²Cm and²⁴⁴Cm for about two years, and for²³⁸Pu,²⁴¹Pu and²⁴¹Am for one year. Concentration ranges are reported. Samples were prepared for counting either directly by evaporation or by chemical separation on BioRad AG 1×4 resin and subsequent electrolysis, and were counted in 20 cm dia Frish grid ionization chambers. Procedures are described. For most actinide nuclides, activity ratios in primary coolant were found to be different from those in worldwide fallout, thus allowing an identification of origin in the case, that actinides should be detected in the vicinity of a nuclear power station.     

Control of the10B/11B isotopic ratio in H3BO3 samples by SSMS

The¹⁰B/¹¹B isotopic ratio in the H₃BO₃ presented in the liquid moderator of a nuclear reactor is used as monitor of the performance of the reactor during the first year of operation. Usually this determination is performed by TIMS due to its high precision. An alternate technique to thermal ionization mass spectrometry /TIMS/ for the determination of the¹⁰B/¹¹B isotopic ratio by SSMS is presented. The modifications introduced in the general analytical method produced very good results. Furthermore, the relative standard deviation was 5% against the typical value of 10–20%.SSMS: spark source mass spectrometry.     

Chemical behaviour and steam transport of nitrogen-13 in BWR primary system

The chemical forms of the¹³N activity in the BWR primary coolant have been determined in four reactors during hydrogen water chemistry tests. Under normal water chemistry conditions, most of the¹³N activity was found in the forms of NO ₂ ^– and NO ₃ ^– in reactor water, and NO ₂ ^– was also found in the steam condensate where the majority of the activity was in NH ₄ ⁺ . On the other hand, under HWC conditions, practically all¹³N activity in both the reactor water and steam condensate was in the NH ₄ ⁺ form. The steam transport of the¹³N activity was significantly enhanced by the addition of hydrogen in the reactor water.     

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.     

A rapid method for the determination of radionuclides in reactor coolant by ion exchange membranes

This paper deals with a rapid method to determine radionuclides in reactor coolant by anion, cation, and anion-cation exchange membranes. A high pressure filtration device was established to simulate the THOR cooling water sampling system by means of several membranes mentioned above. The experimental results indicate that the adsorption efficiency of each membrane for several radionuclides is /1/ >95% with cation exchange membrane for Zn, Co, Na, Mn, Cu, Cs, Ba, La, W etc., /2/ >98% with anion exchange membrane for I, and /3/ <98% with anion-cation exchange membranes for Fe and Cr. The results are obtained using cooling water of Tsing Hua Open-Pool Reactor and the following radionuclides were identified:^99mTc,¹⁴⁰Ba,¹⁴⁰La,⁵¹Cr,¹³¹I,⁵⁸Co,⁶⁰Co,⁵⁴Mn,⁴⁶Sc,⁵⁹Fe,²⁴Na, etc.     

Investigations into the characteristics of irradiated zircaloy cladding material in connection with reactor operation and spent fuel reprocessing

To monitor and improve the performance of nuclear fuel assemblies, the uptake of gases (oxygen, nitrogen and hydrogen) is determined in zircaloy cladding materials. Gas chromatography and/or mass spectrometry combined with hot vacuum fusion were applied. Deviations from the initial concentrations (ca. 5 μg g^?1 H₂, 30 μg g^?1 N₂ and 1200 μg g^?1 O₂) in the “as fabricated” condition, are important in estimating cladding corrosion. The material characteristics of zircaloy are altered substantially by the neutron irradiation and the chemical environment in the reactor coolant, thus the irradiated material must also be studied. For reprocessing safety consideration, the ignition and explosion parameters of unirradiated and irradiated zircaloy dusts were examined. Standard methods, tailored to hot-cell operation, were used to evaluate the minimum ignition temperature of a dust layer on a heated surface at constant temperature, the ignition temperature of a dust cloud, the auto-ignition temperature of a cylindrical dust formation as a function of sample volume, and the explosion pressure and pressure rise in a 20-l spherical chamber. Samples of fines (<100 μm diameter) were characterized by measuring their density and particle-size distribution, and by scanning electron microscopy. For samples of irradiated zircaloy, the ignition temperatures were lower and the explosion pressures and pressure rises higher than for unirradiated zircaloy. These findings can be explained by the different particle-size distribution of irradiated material samples. The increased brittleness of the irradiated material produces more small particles (<20 μm) which favour ignition and explosion.     

Methodology for determination of <Superscript>89</Superscript>Sr and <Superscript>90</Superscript>Sr in radiological emergency: II. Method development and evaluation

In this work a method for the determination of both ⁸⁹Sr and ⁹⁰Sr is presented. The method can potentially be used in radiological emergency and deliver results shortly after an incident. The method development was based on theoretical calculations of potential interferences from other fission products and how these could be discriminated when applying different chemical separation schemes. Validation was done on reactor coolant water containing short-lived fission products, and on a reference material. The results indicate that correct results of ⁸⁹Sr and ⁹⁰Sr can be obtained 4 and 9 days, respectively, after an incident.     

Commercial grade aluminium: A versatile reactor flux monitor

With the advent of Ge(Li) spectrometry, a high standard of purity for neutron flux monitors no longer remains an imperative “must” and becomes rather superfluous. From this standpoint, commercial grade Al was investigated for its suitability as a reactor flux monitor and was found to have a much greater practical utility than most of the monitors reported. Three Al foils and one wire were randomly selected from four different commercial sources, and analysed for their Fe, Ga, Mn and Na contents by neutron activation and high-resolution gamma-spectrometry. While Na was found to have a very heterogeneous distribution, Fe, Mn and Ga concentrations in different splits of each type of Al were consistently uniform within ±2–3%. Eight possible monitor reactions on Al, Mn, Fe and Ga have been recommended as neutron flux integrators for all the 3 components of a reactor spectrum, viz. thermal, epithermal and fast, covering a wide range of flux levels from 10⁷ to 10¹⁴n·cm⁻²·sec⁻¹. The advantages and versatility of commercial grade Al as a pile neutron dosimeter are discussed.     

Effect of pH and phosphate on trapping capacity of various heavy metal ions with ferritin reactor in flowing seawater

We describe a protein reactor consisting of native liver ferritin of Dasyatis akajei (DALF) and a dialysis bag. Our goal was to study a ferritin reactor for its capacity to trap various heavy metal ions (M²⁺) in flowing seawater. The reactor is sensitive and inexpensive and can be operated by nonprofessional technicians. A positive relationship between the number of trapped M²⁺ with the DALF reactor and its concentration in the flowing seawater was observed. Both the pH in the medium and the phosphate content within the ferritin cavity strongly affected trapping capacity. It was found that the ferritin released its phosphate compound directly with a shift in pH without the need for releasing reagent, which differs from the phosphate release characteristics of horse spleen ferritin, as previously described. This behavior evidently makes the trapping capacity with the ferritin reactor weaken, indicating that this trapping capacity is tightly connected to its phosphate compound. Our study shows that a self-regulation ability of the ferritin shell rather than its phosphate compound plays an important role in controlling the rate and capacity of trapping M²⁺. The ferritin reactor was constructed to monitor the contamination level of M²⁺ in flowing seawater. Our preliminary data along with fieldwork indicate that the DALF reactor is an analytical means for effectively monitoring the contamination level of M²⁺ in flowing seawater.     

Determination of hydrogen peroxide concentration in THOR coolant during reactor operation

The concentrations of hydrogen peroxide formed in the coolant due to radiolysis were studied during THOR operation at 1 MW. The relation between doses and hydrogen peroxide formation in a neutron-gamma mixed field was investigated. The initial concentration was 2.3×10^–5 g/ml at the beginning of reactor operation, and then it was increased rather rapidly at the first 9 h. The increasing rate was slowed down till the end of 30 h operation. The maximum concentration of hydrogen peroxide was found to be 4.7×10^–5 g/ml, and its decrease followed the exponential curve.