The FeSe2(cr) solubility determined by solubility experiments of Se co-existing with Fe

To determine the equilibrium constant for ferroselite (FeSe₂(cr)) dissolution reaction, FeSe₂(cr) solubility experiments were performed at 298 ± 1 K from both the over- and under-saturation directions with Fe–Se precipitates that were aged at 348 K. X-ray diffraction (XRD) analysis detected only FeSe₂(cr) as the Se solid phase in the equilibrated precipitates. The E_h values of the equilibrated suspensions ranged from ?188.6 to ?4.9 mV vs. standard hydrogen electrode (SHE) and the pH values ranged from 6.00 to 8.76. Based on the available thermodynamic data, Se₄^2? and Fe²⁺ are thermodynamically stable within this E_h–pH range. Agreement between the solubility data obtained from the over- and under-saturation directions lends credence to the attainment of equilibrium at 298 ± 1 K. The thermodynamic interpretations using the specific ion interaction theory (SIT) model showed that E_h values and the concentrations of Se and Fe are well represented by the 2FeSe₂(cr) solubility reaction (2FeSe₂(cr) ? 2Fe²⁺ + Se₄^2? + 2e^?) with log₁₀K^○ = ?17.09 ± 0.28. The obtained log₁₀K^○ value falls within the uncertainty limits of the log₁₀K^○ value calculated from the available thermodynamic data.     

Boiling and drying accident of high-level liquid waste in a reprocessing plant: Examination of the NO2 and NO generation using the simulated waste

ABSTRACT

NO₂ and NO generated during a boiling and drying accident, which affects the release of volatilized radioactive Ru into the atmosphere, were examined using various samples including simulated high-level liquid waste and a thermogravimetric analyzer. NO₂ and NO in the gas flowing out of the analyzer were measured separately using a NOx analyzer equipped with NO₂ and NO sensors. The samples were heated to 600°C at constant heating rates of mainly 0.2 and 1°C min^?1 that was adopted taking into account the decay heat of high-level liquid waste. It was found that under 180°C some nitrates in the liquid waste mainly separated their nitrate groups as HNO₃ without generating NOx (a mixture of NO and NO₂) and above 300°C the residual HNO₃ in the waste participated in thermal decomposition generating NOx. The generation rates of NO₂ and NO were obtained as a function of time using Arrhenius type equations, and the O₂ rate was derived from these equations using the stoichiometry of the reactions that generate NO₂, NO, and O₂.     

Optimization for removal of ruthenium from nitric acid solution by volatilizing with electrochemical oxidation

Ruthenium is a major fission product element among the platinum group elements (PGEs) in high-level liquid waste (HLLW). Ru tetra-oxide, RuO₄, has high vapor pressure, which is high enough to be run off from its solution even at room temperature. Electrochemical oxidation method, to oxidize nitrosyl ruthenium to the tetra-oxide and then to remove ruthenium from liquid phase to gas phase, was studied to separate Ru from the HLLW. The advantage of this method requires neither additional reagents nor adjustment its valency before the oxidation and disadvantage is necessity of long time for oxidation. In order to improve oxidation rate, we carried out the experiments to clarify the effects following fundamental conditions to the electrochemical oxidation, which are (a) electrolyte temperature, (b) presence of promoter elements, (c) evaporation or reflux of condensed phase, and (d) using or not using of diaphragm at counter electrode. We found the fast oxidation conditions as follows: (1) higher temperature; 95°C, (2) Ce coexistence; 3000 ppm; and (3) usage of a diaphragm for counter electrode. However, evaporation or reflux conditions did not directly affect the electrochemical oxidation efficiency.     

Examination of 131I and 137Cs releases during late phase of Fukushima Daiichi NPP accident by using 131I/137Cs ratio of source terms evaluated reversely by WSPEEDI code with environmental monitoring data

To investigate what happened in reality during the Fukushima Daiichi Nuclear Power Plant accident, the phenomena within reactor pressure vessel and the discussion of ties with the environmental monitoring measurement are very important. However, the previous study that treats phenomena of the both has not necessarily advanced up to the present time. The source terms predicted by simulation codes such as MELCOR has not yet been consistent with the reverse estimation by WSPEEDI code using environmental measurement data. This study investigated ¹³¹I and ¹³⁷Cs release behaviors during the late phase of the accident to contribute to such examination using the ¹³¹I/¹³⁷Cs ratio of the new source terms predicted by Katata. The ¹³¹I release by the gas–liquid partition from the contaminated water in the 1F2 and 1F3 reactor buildings which was pointed out in the previous study was reevaluated using the new source terms. In addition, paying attention to the similarity of the core conditions between the Fukushima accident and the Phébus FPT3 experiment using the B₄C control rods, the release of organic iodine (CH₃I) during the 1F3 suppression pool venting, formation of CsBO₂ and its release behavior were examined which have not yet been sufficiently studied so far.     

Determination of 79Se and 135Cs in Spent Nuclear Fuel for Inventory Estimation of High-Level Radioactive Wastes

⁷⁹Se and ¹³⁵Cs are long-lived fission products and are found in high-level radioactive waste (HLW). The estimation of their inventories in HLW is essential for the safety assessment of geological disposal, owing to their mobility in the strata. In this study, the amounts of ⁷⁹Se and ¹³⁵Cs in a spent nuclear fuel solution were measured. About 5 g of irradiated UO₂ fuel discharged from a commercial Japanese pressurized water reactor (PWR) with a burn-up of 44.9 GWd/t was sampled and dissolved with 50mL of 4M nitric acid in a hot cell for 2 h. After Se and Cs were chemically separated, the amounts of ⁷⁹Se and ¹³⁵Cs in the spent nuclear fuel solution were measured by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS). The amounts of ⁷⁹Se and 135Cs were 5:2 ± 1:5 and 447 ± 40 g/MTU, respectively. The results presented in this study, which are the first postirradiation experimental data in Japan, showed good agreement with those obtained by the ORIGEN2 code using the data library of JENDL-3.3.     

A new value for the half-life of Be by Heavy-Ion Elastic Recoil Detection and liquid scintillation counting

The importance of ¹⁰Be in different applications of accelerator mass spectrometry (AMS) is well-known. In this context the half-life of ¹⁰Be has a crucial impact, and an accurate and precise determination of the half-life is a prerequisite for many of the applications of ¹⁰Be in cosmic-ray and earth science research. Recently, the value of the ¹⁰Be half-life has been the centre of much debate. In order to overcome uncertainties inherent in previous determinations, we introduced a new method of high accuracy and precision. An aliquot of our highly enriched ¹⁰Be master solution was serially diluted with increasing well-known masses of ⁹Be. We then determined the initial ¹⁰Be concentration by least square fit to the series of measurements of the resultant ¹⁰Be/⁹Be ratio. In order to minimize uncertainties because of mass bias which plague other low-energy mass spectrometric methods, we used for the first time Heavy-Ion Elastic Recoil Detection (HI-ERD) for the determination of the ¹⁰Be/⁹Be isotopic ratios, a technique which does not suffer from difficult to control mass fractionation. The specific activity of the master solution was measured by means of accurate liquid scintillation counting (LSC). The resultant combination of the ¹⁰Be concentration and activity yields a ¹⁰Be half-life of T_1/2 = 1.388 ± 0.018 (1 s, 1.30%) Ma. In a parallel but independent study (Chmeleff et al. [11]), found a value of 1.386 ± 0.016 (1.15%) Ma. Our recommended weighted mean and mean standard error for the new value for ¹⁰Be half-life based on these two independent measurements is 1.387 ± 0.012 (0.87%) Ma.     

Atom probe characterization of copper solubility in the Midland weld after neutron irradiation and thermal annealing

An atom probe field ion microscopy characterization has been performed to determine the copper matrix concentration in a submerged arc beltline weld of the Midland Unit 1 pressurized water reactor after four conditions: unirradiated, unirradiated and annealed for 168 h at 454°C, neutron-irradiated in a test reactor to a fluence of 1.1×10²³ n m⁻² (E>1 MeV) at a temperature of 288°C, and neutron-irradiated and annealed for 168 h at 454°C. Atom probe analysis of the unirradiated material revealed a substantial depletion of the copper in the matrix to 0.119±0.007 at.% Cu from the bulk value of between 0.18 and 0.28 at.% Cu. Annealing the unirradiated material produced intragranular copper-enriched precipitates and reduced the matrix copper level by 25% to 0.088±0.012 at.% Cu. Neutron irradiation also produced copper-enriched precipitates and reduced the matrix copper level by almost 50% over the stress relieved material to 0.058±0.008 at.% Cu. Annealing the neutron-irradiated material reduced the matrix copper level further to 0.050±0.010 at.% Cu. These results indicate that the annealing treatment coarsens the copper-enriched precipitates produced during neutron irradiation with a slight decrease in the matrix copper content.     

Frictional Pressure Drop for NaK-N2 Two-Phase Flow in Rectangular Cross Section Channel of Large Aspect Ratio

In this paper, the frictional pressure drop in an isothermal liquid metal-gas two-phase flow through a rectangular channel with large width-to-height ratio is treated semiempirically for a NaK-N₂ two-phase flow system.

The frictional pressure drop in the two-phase flow is compared with the following two reference values :

1. The frictional pressure drop in the liquid flowing alone in single phase with the same velocity as that of the liquid in the two-phase mixture.

2. The frictional pressure drop in the liquid flowing alone in single phase with the same mass flow rate as that of the liquid in the two-phase mixture.

The comparison with the former reference value is necessary for the prediction of friction loss in a liquid metal MHD generator channel whose medium would be two-phase mixture.

The semiempirical analysis was performed assuming the two-phase mixture to be a continuous medium with its properties, e.g. viscosity and density, defined by void fraction and the velocity determined by the total mass flow rate.

In the region of low slip and density ratio ρ_g/ρ_l the frictional pressure drop in the two-phase flow appeared to be smaller than that due to the liquid flowing alone with the same velocity as that of the liquid in the two-phase flow.

The experiments have been undertaken with the NaK-N₂ two-phase mixture flowing through a rectangular channel (4 × 60 mm²).

Data were taken over the following parameter range:

NaK velocity: 5～30 m/sec, Void fraction: 0～70%

Density ratio: 0.006～0.013, Quality: 0.07～1.10%.     

Atomistic simulations on the interfacial interaction of metallic fuel and structural materials in SFRs - molecular dynamics model for Pu-Fe system

Interaction between metallic fuel and steel structures is one of the predominant phenomena in the progress of core disruptive accidents of Sodium-cooled fast reactor. In this study, the atomic diffusion across the interface between Pu and Fe was investigated by using molecular dynamics. The simulation was performed by using Modified Embedded Atom Method (MEAM). The interactions between plutonium and iron atoms were calculated by using the newly developed potential model determined so as to reproduce the material properties of PuFe₂ and Pu₆Fe. The material properties of the compounds predicted with the developed potential were in good agreement with the referenced data. The dissolution or melting at the interface between solid Fe and solid or liquid Pu were simulated by contacting semi-infinite slabs (or liquid layer) of them. Dissolution was observed for all the tested temperature conditions from 800 K up to 1700 K. The melting at the interface was also observed on the interface between solid Fe and PuFe₂ slabs at the temperature approximately 100 K below the melting temperature of PuFe₂ obtained based on the present model.     

Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water

The low-frequency corrosion fatigue (CF) crack growth behaviour of different low-alloy reactor pressure vessel steels was characterized under simulated boiling water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in the temperature range of 240-288 °C with different loading parameters at different electrochemical corrosion potentials (ECPs). Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In this paper the effect of ECP on the CF crack growth behaviour is discussed and compared with the crack growth model of General Electric (GE). The ECP mainly affected the transition from fast (‘high-sulphur’) to slow (‘low-sulphur’) CF crack growth, which appeared as critical frequencies ν_crit = f(ΔK, R, ECP) and ΔK-thresholds ΔK_EAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dt_Air,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔK_EAC-thresholds were shifted to lower values with increasing ECP. The CF crack growth rates of all materials were conservatively covered by the ‘high-sulphur’ CF line of the GE-model for all investigated temperatures and frequencies. Under most system conditions, the model seems to reasonably well predict the experimentally observed parameter trends. Only under highly oxidizing conditions (ECP ? 0 mV_SHE) and slow strain rates/low loading frequencies the GE-model does not conservatively cover the experimentally gathered crack growth rate data. Based on the GE-model and the observed cracking behaviour a simple time-domain superposition-model could be used to develop improved reference CF crack growth curves for codes.     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.