In order to rationalize maintenance for nuclear power plants, it is necessary to develop optimize maintenance plan by considering characteristics of each plant. In sodium-cooled fast reactor, there are constraints on inspections due to the specialty of handling sodium equipment, that is one of the important points when considering rationalization of maintenance. To solve this problem, we proposed a basic concept of maintenance optimization scheme that is a design support tool in order to develop maintenance strategy, based on “system based code (SBC)”. SBC is a concept to optimize the reliability of a nuclear power plant by consideration of all related technical requirements. “ASME Code Case N-875” and “ASME Boiler and Pressure Vessel Code, Section XI, Division 2 (RIM)” based on system based code were already developed as standards for inspection. One of the proposed scheme goals is to make a concrete way of necessary assessment method. Another is to provide several combinations of design and maintenance, and information for owner in order to choose the acceptable combination. In the beginning, we are working to develop the scheme that can be applied to sodium fast reactor as the main concept of next generation reactor. In this context, primary heat transfer system (PHTS) piping of fast reactor was evaluated by the scheme. This piping was chosen because it is major significant component and the inspection have constraint conditions that need preparation work. As a result, design candidate (e.g. single and double wall piping) and inspection candidate (e.g. ultrasonic testing and continues leakage monitoring) combinations along with benefit of each cases were provided.

Liquid sodium to be described by sodium here after with suspended nanoparticles is expected to mitigate sodium-water reaction, because sodium with suspended nanoparticles will suppress the degree of chemical activity by the atomic interaction between nanoparticles and sodium. In order to investigate this interaction from the view point of physical property, we tried to measure the surface tension of sodium with suspended nanoparticles in a static and dynamic state. For the measurement of the surface tension in a static state, we employed the pendant drop method. The measurements were done in the globe box keeping inert gas atmosphere, which was required for the experiment using sodium to get rid of an easy chemical reaction of sodium with oxygen and moisture. We succeeded to make pendant drops of sodium and sodium with suspended nanoparticles between 200°C and 500°C. On the other hand, to evaluate the surface tension in a dynamic state, we employed the method applying the principle of the Kelvin-Helmholtz instability. The measurements were done in the globe box used in the static experiment. Sodium or sodium with suspended nanoparticles throwing a nozzle was ejected into a paraffin pool. Its fragments were made due to the Kelvin-Helmholtz instability. The fragments were retrieved from a paraffin pool and we measured fragment size distribution. We carried out the experiments at temperature of 150°C and 200°C, and the ejection volume rate of 3.0L/min and 6.0L/min. It was found that droplet sizes of sodium with suspended nanoparticles were increased as compared with those of sodium without nanoparticles. This result physically must have mean that sodium with suspended nanoparticles has larger surface tension than that of sodium without nanoparticles in the present condition. Therefore, we could consider the effect of the atomic interaction might mitigate sodium-water reaction.

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 ( P rogram SY stem for C orrosion H azard E valuation) 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.

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.

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 Na 2 I 2 in the cover gas.

Reaction behavior of carbon dioxide (CO2) with a liquid sodium pool was experimentally investigated to understand the consequences of boundary tube failure in a sodium-CO2 heat exchanger. In this study, two kinds of experiments were carried out to investigate the reaction behavior. In one experiment, about 1–5g of liquid sodium pool were poured into flowing CO2 to obtain the information mainly about the thermo-chemical conditions to initiate the reaction and the chemical constituents of reaction products. During the experiment, visual observation was made using video-camera and the temperature change of the sodium pool and near the surface was measured by thermocouples. The experimental parameters were the sodium pool diameter, the initial temperature of sodium and CO2, the CO2 flow direction against pool surface, and the initial moisture concentration in CO2. The solid products of sodium-CO2 reaction were sampled and analyzed by X-ray diffraction (XRD), Energy Dispersion X-ray analysis (EDX), Total Organic Carbon analysis (TOC), and chemical analysis. The reaction gas products were also sampled and analyzed by gas chromatography. In the other experiment, CO2 was injected into about 200g of liquid sodium pool to simulate the boundary failure in the sodium-CO2 heat exchanger. The CO2 was fed through a helical coil-type tube dipped into the pool to adjust the temperature to the sodium pool temperature, and injected upward into the pool from a pool bottom using a nozzle attached at the end-side of the tube. The experimental parameters were the initial temperature of sodium, the diameter of the nozzle, the flow rate and the injection time of CO2. The temperature change of sodium pool and the cover gas was measured by thermocouples during the experiment, and the reaction products were sampled and analyzed by the same manner as in the former experiments after the experiment. From these experiments, it became clear that the exothermic reaction occurred above a threshold temperature, and useful and indispensable information such as the resulting temperature and pressure rise and the behavior of solid reaction products in the pool was obtained to evaluate the consequences of boundary tube failure incident in a sodium-CO2 heat exchanger.

In the framework of PSI’s FAST code system, the TRACE thermal-hydraulics code is being extended for representation of sodium two-phase flow. As the currently available version (v.5) is limited to the simulation of only single-phase sodium flow, its applicability range is not enough to study the behavior of a Sodium-cooled Fast Reactor (SFR) during a transient in which boiling is anticipated. The work reported here concerns the extension of the two-fluid models, which are available in TRACE for steam-water, to sodium two-phase flow simulation. The conventional correlations for ordinary gas-liquid flows are used as basis, with optional correlations specific to liquid metal when necessary. A number of new models for representation of the constitutive equations specific to sodium, with a particular emphasis on the interfacial transfer mechanisms, have been implemented and compared with the original closure models. As a first application, the extended TRACE code has been used to model experiments that simulate a loss-of-flow (LOF) accident in a SFR. The comparison of the computed results, with both the experimental data and SIMMER-III code predictions, has enabled validation of the capability of the modified TRACE code to predict sodium boiling onset, flow regimes, dryout, flow reversal, etc. The performed study is a first-of-a-kind application of the TRACE code to two-phase sodium flow. Other integral experiments are planned to be simulated to further develop and validate the two-phase sodium flow methodology.

Sodium cooled fast breeder reactors constitute the second stage of India’s three-stage nuclear energy programme, for effective utilization of the country’s limited reserves of natural uranium and exploitation of its large reserves of thorium. The Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, is a sodium cooled, loop type fast reactor. Its main aim is to provide experience in fast reactor operation and large scale sodium handling and to serve as a test bed for irradiation of fast reactor fuels and materials. FBTR was built on the lines of the French Rapsodie-Fortissimo reactor, with modifications to make it a generating plant. FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. The steam water system mainly consists of a once-through steam generator, which produces super heated steam at a pressure of 125 bars and temperature of 480° C, feed water system and condensate system. The steam produced is supplied to a condensing turbine coupled to an alternator. The reactor achieved first criticality in Oct 85 with a small core of 22 fuel subassemblies (SA) having a unique carbide fuel rich in Pu. This fuel (called MK-I) was developed and made in India and has a composition of 70% PuC-30% UC. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. In the light of the excellent performance of the carbide fuel, which has endured a burn-up of 155 GWd/t without any clad failure, the core has been gradually expanded by the addition of mark II (55% PuC-45% UC) and MOX (44% PuO 2 −56% UO 2 ) fuel SA to compensate for the burn-up reactivity loss, and the reactor power has been progressively raised from 10.6 MWt to a maximum of 17.4 MWt. Over the years, several safety related experiments have been conducted. These include natural convection tests and experiments to validate the Failed Fuel Detection System. The challenges faced include a major fuel handling incident, primary sodium leak and reactivity transients. Two major modifications were carried out — one on the Steam Generator Leak Detection System and the other in the steam-water circuit. These helped in improving the campaign availability from less than 50% to more than 90%. The main component limiting the life of reactor is the grid plate supporting the core. The fast flux at the grid plate was measured using Np foils. The residual life of the grid plate has been estimated as 11 effective full power years. The paper presents operating experience of the reactor, performance of the carbide fuel, safety related experiments done in the reactor, various challenges faced, various modifications carried out to improve system reliability and availability and residual life assessment of the reactor.

Liquid sodium is used as the coolant of the fast breeder reactor (FBR). A weak point of sodium is a high chemical reactivity with water or oxygen. So an idea of chemical reactivity suppression of liquid sodium itself is proposed. The idea is that nano-meter size particles (hereafter called nano-particles) are dispersed in liquid sodium, and an atomic interaction which is generated between nano-particle and sodium atoms is applied to suppress the chemical reactivity. We call sodium that has dispersed the nano-particle a Nano-fluid. Three key technologies which are the trial manufacture of Nano-fluid, the reaction property of the Nano-fluid and applicability of Nano-fluid to FBR Plant have been carried out to develop the Nano-fluid.

In sodium-cooled fast reactors, it is very important to evaluate sodium-water reaction from the viewpoint of safety. If once it happens in the steam generator, it is desirable its scale is small and its reaction is mild. The sodium dispersing nano particles is expected to mitigate the sodium-water reaction. Nano particles is titan particles whose diameter is tens nanometer and dispersing rate in sodium is several percent. In order to find out the effect clearly, it is essential to evaluate the reaction behavior and characteristics of physical properties closely. For this purpose, we tried to measure the surface tension of sodium dispersing nano particles as an important part of foundational physical property. We had employed the pendant drop method for the measurement, with which we can measure surface tension from the shape of pendant drop. This method is suitable for our measurement because nano sample is very limited. The measurement was done in a sealed box keeping inert gas atmosphere, which is required for sodium experiment because of an easy chemical reaction of sodium with oxygen and moisture. At first, we did a series of experiments for pure sodium. We succeeded to make pendant drops between 150°C and 350°C. For the measurement, we selected only the drops in clear mirror surface with axial symmetry. The scatter range of obtained results is slightly larger than that of a recommended equation. We estimated that it was caused due to a weak wet condition on the edge face of nozzle. Then we measured the surface tension of the sodium dispersing nano particles under the same condition. The obtained results almost show values larger than those of pure sodium. The dispersion rate of nano particles has an uncertainty in the present study. However even in giving the worst evaluation, the measured surface tensions for the sodium dispersing nano particles show larger values than those for the pure sodium.

Experiments of density wave instability in a liquid sodium boiling two-phase flow experiments in an annulus were carried out in the following parameters range: heat flux from 80kW/m 2 to 976kW/m 2 , inlet subcooling from 25.6°C to 226.8°C, mass flow rate from 7.92kg/h to 68.9kg/h, system pressure from 2600Pa to 0.12Mpa. Not only the mechanism of the instability, critical conditions and oscillation period, but also the effects of pressure, mass flow rate and inlet subcooling on the density wave instability were explored experimentally and theoretically. From the experimental data, it was found that the lower the inlet temperature was, the higher the system pressure and the mass flow rate that could result in a more stable boiling two-phase flow were. A correlation for the density wave instability was obtained on from the dimensional analysis for the conservation equations of mass, momentum and energy.

The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m 3 (86,000 gallons) of sodium and the secondary system contained 50 m 3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P&IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.

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 dm 3 /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.