Shinya Miyahara

Equilibrium Evaporation Behavior of Polonium and Its Homologue Tellurium in Liquid Lead-Bismuth Eutectic

Experimental study using the transpiration method investigates equilibrium evaporation behavior of radionuclide polonium (210Po) generated and accumulated in liquid lead-bismuth eutectic (LBE) cooled nuclear systems. The experiment consists of two series of tests: preliminary evaporation tests for homologue element tellurium (Te) in LBE, and evaporation tests for 210Po-accumulated LBE in which test specimens are prepared by neutron irradiation. The evaporation tests of Te in LBE provide the suggestion that Te exists in a chemical form of PbTe as well as the information for confirming the validity of technique and conditions of Po test. From the evaporation tests of 210Po in LBE, we obtain fundamental data and empirical equations such as 210Po vapor concentration in the gas phase, 210Po partial vapor pressure, thermodynamic activity coefficients, and gas-liquid equilibrium partition coefficients of 210Po in LBE in the temperature range from 450 to 750°C. Additionally, radioactivity concentration of 210Po and 210mBi vapor in a cover gas region of a typical LBE-cooled nuclear system is specifically estimated based on the obtained experimental results, and the importance of 210Po evaporation behavior is quantitatively demonstrated.

3D RFEC simulations for the in-service inspection of steam generator tubes in fast breeder reactors

Three-dimensional numeric simulations, based on the finite element model, are conducted to simulate the In-Service Inspection of magnetic steam generator tubes of a Fast Breeder Reactor using remote field eddy current probes. The influen ce of sodium in the gap SP-tube and outer defect detectability is computed for a large SP model with multiple SG tubes using the 3D-RFECT code.

Reactive Wetting of Metallic Plated Steels by Liquid Sodium

Sodium wetting experiments were performed to investigate the reactive wetting of metallic plating materials by liquid sodium at 250°C for the ultrasonic sensor of the under-sodium viewer. SUS304 stainless steel specimens were electrolytically plated with four metallic materials (nickel, palladium, gold, and indium) that have different solubilities in sodium, and the spreading velocity of sodium droplets on the metallic plated specimens was measured. It was confirmed that the spreading velocity increased as the solubility increased, and the constant α on the spreading velocity on the plated specimens was unique for the plating materials and was proportional to the logarithm of the solubility of the plating materials. Furthermore, it is considered possible to select plating materials based on solubility from the result of this study.

Prediction of Radioactive Corrosion Product Transfer in Primary Systems of Japanese Prototype Fast Breeder Reactor MONJU

Radioactive corrosion products (CP) are main cause of personal radiation exposure during maintenance with no breached fuel in FBR plants. CP is produced in the core region by activation of fuel cladding and sub-assembly wrappers, and they are transported to the primary circuit with sodium flow and deposited on the wall of the primary piping and components. In order to establish the techniques of radiation dose reduction for of personnel, program system for corrosion hazard evaluation code “PSYCHE” has been was developed. The PSYCHE code is based on the solution-precipitation model. The density of each deposited CP and dose rate of primary coolant system in MONJU was estimated by using the PSYCHE and QAD-CG code. From the calculation for MONJU, it was predicted that deposition inventory of 54Mn would be about ten times more than that of 60Co in 20 years. In particularly deposition density; 54Mn would deposit mainly in the primary pump and cold-leg. On the contrary, 60Co depositions to not only the primary pump but also the Hot-leg. In the result, the future values of dose rate around primary piping system were estimated to be saturated at 2∼3 mSv/h. The calculation result of CP suggests that the analysis code and model make it possible to differentiate behavior of these nuclear species, definitely. Using the analysis code, the prospects for the future operations of MONJU could be established. Moreover, it is necessary to examination of improvement that the innovative calculation procedure and physics model, for applied to next generation LMFBR such as JSFR.Copyright

Experimental Measurements of Eddy Current Signal From SG Tubes of Fast Breeder Reactor Covered by a Thin Sodium Layer Using a SG Mock-Up

In Fast Breeder Reactors (FBR) which are sodium cooled, the steam generator (SG) heat exchanger tubes separate the low pressure sodium flowing in the SG vessel with the high pressure water-steam in tubes. During In-Service Inspection (ISI), sodium is first drained and then SG tubes are cooled down to the room temperature. After sodium draining, due to the high temperature (more than 500 °C), sodium adheres to SG tubes and structures around (SG support plates, welds) in a thin layer, filling eventually the gaps between SG support plates and tubes. During ISI, SG tubes are inspected for cracks and corrosions using differential eddy currents (EC) probes. Due to the high electrical conductivity of sodium adhering to the outer SG tube surface, the eddy current testing (ECT) signal modifies, in accord with sodium layer thickness or sodium deposits located on the outer SG tube surface. The sodium wetting properties depends on several factors as: material surface, temperature and sodium wetting time. The effect of sodium adhering to the outer SG tube on ECT signals were measured using a small mock-up tank (2 m high and 0.7 m in diameter) in which were introduced two SG tubes similar with the ones used in the Monju FBR (one tube is ferromagnetic and made of 2.25Cr–1Mo alloy, while the other one is made of SUS321 and is austenitic). Defects, SG support plates (on both helical and straight part of the tube) and welds were added to tubes and the ECT signal was measured before and after sodium draining. Variations in the sodium layer thickness and consequently its effect on ECT signals were measured by filling and draining the tank three times in order to recreate each time new layers of sodium. The paper describes the experimental conditions and the ECT results for both types of SG tubes by comparing the defects, SG support plates and weld signals before and after draining of sodium. Additionally, sodium structures were examined visually using a VideoScope camera, confirming the recorded ECT signals. The paper also presents details about sodium layer thickness measurements in several parts of SG tubes (near defect, SP, weld, bend, helical tube, straight tube) by scratching and collecting the sodium on a small area of 20mm×20mm. The volume of sodium drops is also estimated. The measurement results showed that there are significant differences in the sodium layer thickness depending on the SG tube material.© 2009 ASME

EDDY CURRENT SIMULATIONS AND MEASUREMENTS OF SODIUM EFFECT FOR MAGNETIC AND NON‐MAGNETIC STEAM GENERATOR TUBES OF FBR

In fast breeder reactor (FBR), the steam generator (SG) tube wall is the only barrier between water steam and sodium flow. Eddy current signal (ECT) from outer tube defect is modified by both SG support plates (SP) as well as by sodium layer and unknown sodium drops located on the outer SG tube surface. In the present paper, ECT finite element simulations are conducted to evaluate sodium structures ECT noise and variations of defect and tube support plate signal in the presence of a thin layer covering the SG tube surface. Numerical simulations are validated and calibrated with experimental measurements of artificial outer defect for both magnetic and non‐magnetic SG tubes in the absence or presence of sodium covering the outer surface of SG tubes. The papers presents also details about measurements of sodium structures (drops, layer) formed on the outer SG tube surface when these are soaked in a test tank filled with sodium at high temperatures (500° C) up to two hours.

Reaction, Transport and Settling Behavior of Lead-Bismuth Eutectic in Flowing Liquid Sodium

The experimental study has been carried out to investigate reaction, transport and settling behavior of lead-bismuth eutectic (LBE) in flowing liquid sodium. In the test, 168g of LBE were poured into flowing sodium from the top of a vertical-type sodium loop which contained 23.2 kg of sodium. The initial temperature of LBE and sodium was 673K. The flow rate and the maximum velocity of sodium in the loop were controlled and measured at 20 dm3 /min and 1 m/sec, respectively, using an electro-magnetic pump and an electro-magnetic flow meter. The sodium loop has a settling chamber at the lower part to investigate the concentration decrease behavior of solid particle reaction products in the sodium due to the settling effect. The concentration was measured by sodium sampling from the 11 positions of the loop during the experiment and its post-test chemical analysis. The temperature changes at the various parts of the loop were also measured during the experiment by thermo-couples attached on the outer surface of the loop. Ultrasonic detectors were attached on the outer surface of the loop below the position of a LBE pour nozzle to demonstrate the utility as a leak detector.Copyright

A study on self-terminating behavior of sodium–concrete reaction

ABSTRACT A sodium–concrete reaction (SCR) is one of the important phenomena to cause the structural concrete ablation and the release of hydrogen (H2) gas in the sodium (Na) leak accident. In this study, the long-time SCR test had been carried out to investigate the self-termination mechanism under the condition to keep the temperature of Na on the concrete more than the reaction threshold temperature during 24 hours. The test results showed the SCR terminated by itself even if enough amount of Na remained on the concrete. In addition, quantitative data were collected on the SCR terminating behavior such as the temperature, the concrete ablation depth, the H2 generation behavior and the concentration profiles of Na, silicon (Si), aluminum (Al) and calcium (Ca) in the reaction products after the test. In the concentration profiles, the calculation by the sedimentation diffusion model of the steady state was comparable with the experimental results. Though the reaction products were suspended by H2 bubbling and Na ablated the concrete surface with the high H2 generation rate, the reaction products gradually settled down with decreasing of the H2 generation rate. Therefore, the Na concentration decreases at the reaction front with time and the SCR terminates of itself.

Development of Evaluation Methods for the Transfer Behavior of Corrosion Product in the Primary Cooling System of Fast Breeder Reactor

Radioactive corrosion products (CPs) are a main cause of personal radiation exposure during maintenance with no breached fuel in sodium-cooled fast breeder reactor (FBR) plants. In order to establish techniques of radiation dose estimation for workers in radiation-controlled areas of the fast breeder reactor, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. For the inspection of the analysis model using PSYCHE code, we surveyed past reports on CP deposits in Japan Experimental Fast Reactor JOYO and sodium test loops. The SEM images of the external surface of the irradiated fuel cladding in JOYO show surface deposition of particles containing significant volumes of CP species. The CP particle deposition model (Particle model) was developed from the phenomenological considerations of the observation results for JOYO. We add the Particle model to the conventional PSYCHE analytical model. Moreover, the conventional PSYCHE code does not consider the sodium flow of the coolant. The non-consideration of the sodium flow is a cause of reduction in calculation precision. To counter this problem we built the system that PSYCHE code linked NETFLOW++ code. This code is the one-dimensional network code which can simulate the plant transients of various types of nuclear reactors. In this study, we performed calculations of CP transfer behavior in JOYO using an improved PSYCHE code. The calculation results are consistent with the measured results for actual components in JOYO. The C/E (Calculated / Experimentally observed) value was improved by introduction of the Particle model and considered of the sodium flow.Copyright

Equilibrium Partition Coefficients of Cesium and Iodine Between Sodium Pool and the Inert Cover Gas

Equilibrium partition coefficients were experimentally measured for volatile fission products of cesium and iodine between liquid sodium pool and the inert cover gas. In the experiments, the “transpiration method” was utilized in which the saturation vapor of sodium with cesium and iodine vapor in an isothermal evaporation pot was transported by inert carrier gas and trapped by filters outside the pot. The objectives of the experiments are to: a) Obtain the equilibrium partition coefficients of cesium and iodine at high temperature between 600 and 850 deg-C, and b) Study the dependence of the partition coefficients upon the concentration in the sodium pool. From the results of previous work and this study, the following empirical equations between the partition coefficients of cesium and iodine and the sodium pool temperature could be obtained: log Kd(Cs) = 2173/T − 1.0487 (from 450 to 850 deg-C)log Kd(I) = −215/T − 0.271 (from 450 to 850 deg-C) These equations are consistent with Castleman’s theoretical equations. The partition coefficients of cesium measured at five different points of mole concentration in the pool were almost consistent with the theoretical values and decreased with the increase in the concentration. On the other hand, the measured partition coefficients of iodine increased with the increase in the concentration in the pool and this tendency was incompatible with the theoretical consideration. The reason of this discrepancy might be attributed to the formation of Na2 I2 in the cover gas.© 2010 ASME