Fire suppression systems for transuranic (TRU) waste facilities are designed to minimize radioactive material release to the public and to facility employees in the event of a fire. Currently, facilities with Department of Transportation (DOT) 7A drums filled with TRU waste follow guidelines that assume a fraction of the drums experience lid ejection in case of a fire. This lid loss is assumed to result in significant TRU waste material from the drum experiencing an unconfined burn during the fire, and fire suppression systems are thus designed to respond and mitigate potential radioactive material release. However, recent preliminary tests where the standard lid filters of 7A drums were replaced with a UT-9424S filter suggest that the drums could retain their lid if equipped with this filter. The retention of the drum lid could thus result in a very different airborne release fraction (ARF) of a 7A drum’s contents when exposed to a pool fire than what is assumed in current safety basis documents. This potentially different ARF is currently unknown because, while studies have been performed in the past to quantify ARF for 7A drums in a fire, no comprehensive measurements have been performed for drums equipped with a UT-9424S filter. If the ARF is lower than what is currently assumed, it could change the way TRU waste facilities operate. Sandia National Laboratories has thus developed a set of tests and techniques to help determine an ARF value for 7A drums filled with TRU waste and equipped with a UT-9424S filter when exposed to the hypothetical accident conditions (HAC) of a 30-minute hydrocarbon pool fire. In this multi-phase test series, SNL has accomplished the following: (1) performed a thermogravimetric analysis (TGA) on various combustible materials typically found in 7A drums in order to identify a conservative load for 7A drums in a pool fire; (2) performed a 30-minute pool fire test to (a) determine if lid ejection is possible under extreme conditions despite the UT-9424S filter, and (b) to measure key parameters in order to replicate the fire environment using a radiant heat setup; and (3) designed a radiant heat setup to demonstrate capability of reproducing the fire environment with a system that would facilitate measurements of ARF. This manuscript thus discusses the techniques, approach, and unique capabilities SNL has developed to help determine an ARF value for DOT 7A drums exposed to a 30-minute fully engulfing pool fire while equipped with a UT-9424S filter on the drum lid.

All of the boiling water reactors in Japan were required to install filtered containment venting system (FCVS) for restart after the accident of Fukushima Daiichi Nuclear Power Plant. FCVS is composed of alkaline water and a metal fiber filter (MFF). Alkaline water has the function of reducing aerosol and inorganic iodine (I2). MFF is constituted for the purpose of removing the aerosol which was not able to be removed with alkaline water. With the above system, the released aerosol can be removed with DF1000 and inorganic iodine DF100. On the other hand, organic iodine that cannot be removed by alkaline water and MFF is removed by silver zeolite added downstream of the FCVS because of its high scattering property. Silver zeolite has particle of silver in the pores of the zeolite. Organic iodine is removed by chemical bonding with the silver. In the current system, the silver zeolite is DF50. Hitachi has developed the organic iodine removal system focusing on removing liquid chemicals for the purpose of improving the performance andmaintain ability, and reducing the price. In this presentation, we report the ionic liquid (IL) that has high heat resistance, high radiation resistance and high gas adsorption among liquid chemicals. In the small-scale tests, we found that IL has a higher performance than DF250, and improved maintainability by liquid property that is able to discharge with scrubber water at the same time, and have got the prospect of reducing manufacturing costs.

SG (steam generator) is one of the most important equipment in fast reactors, the experience in design and operation of fast reactor worldwide show that failures of SG occurred frequently and often caused serious consequences, therefore it’s necessary to conduct reliability analysis on SG in design phase. FMEA (Failure Mode Effect Analysis) is used to identify all potential failure modes and filter out main failure modes. Then, qualitative analysis and quantitative calculation are carried out to evaluate main failure modes. Next, reliability of SG can be obtained by conducting Latin Hypercube Sampling. Analysis results show that the leakage probability of SG in 20 years is 0.130 219, and the most sensitive factor is the quality of weld in the junction of tubes and tube plate, and the SG meet its reliability requirement.

Experience feedback refers to the timely information collection, transformation, analysis, processing and summary, when some good experience or occasional problems appeared during the manufacturing process. In the manufacturing process of nuclear fuel, CNNC JianZhong Nuclear Fuel Assembly Co., Ltd (CJNF) established a comprehensive experience feedback system, and consolidated experience feedback processing flow. Using classification and gradation to collect, filter, organize and use information. CJNF feeds back some quality problems in fuel assembly manufacture, through the analysis of causes and the implementation of measures to avoid the occurrence of similar problems. Meantime, feeding back and sharing good practical experience in manufacturing and management process, it is benefit to pass on experience and learn from each other. The experience feedback of CJNF is the solid foundation of quality management system’s operation and improvement.

The highly turbulent flow inside a pressurized water reactor makes unpractical the use of scale resolving simulations, due to the large number of space and time turbulent structures. The high computational cost associated with typical large eddies simulations or direct numerical simulations techniques is unsuitable due to the large spatiotemporal resolution required. Partially averaged Navier-Stokes turbulence model is presented as bridging model between Reynolds averaged Navier-Stokes equations and direct numerical simulations. As filtered representation of the Navier-Stokes equations, the model is able to continuously shift its energy-based filter, inside the turbulence spectrum, being able to resolve the turbulent scales of interest. The choice of energy based cut-off filters gives the chance to directly impose the degree of needed resolution, where the most important large scales unsteadiness are resolved at minimal computational expenses. The partially averaged Navier-Stokes modelling approach has been tested for a Reynolds number of 14,000, inside a 5 × 5 fuel bundle, with a single spacer grid and split-type mixing vanes. Four different filters have been tested, whose resolution ranged from Reynolds averaged Navier-Stokes and large eddy simulation. A comparison with large eddy simulation will be presented. The results show that the partially averaged Navier-Stokes modeling produces results comparable to those of large eddy simulation when the appropriate cut-off energy filter is chosen. The turbulence models results will be compared with the available particle image velocimetry experimental data.

During a severe accident of a nuclear reactor, radioactive aerosols may be released from degraded nuclear fuels. Pool scrubbing is one of the efficient filters with a high aerosol removal efficiency, in other words a high decontamination factor (DF). Because of its high performance, many pool scrubbing experiments have been performed and several pool scrubbing models have been proposed. In the existing pool scrubbing experiments, an experimental condition of aerosol number concentration was seldom taken into account. It is probably because DF is assumed to be independent of aerosol number concentration, at least, in the concentration where aerosol coagulation is limited. The existing pool scrubbing models also follow this assumption. In order to verify this assumption, we performed a pool scrubbing experiment with different aerosol number concentrations under the same boundary conditions. The test section is a transparent polycarbonate pipe with an inner diameter of 0.2 m. 0.5 μm SiO 2 particles were used as aerosols. As a result, DF was increasing as decreasing the aerosol number concentration. In order to ensure a reliability of this result, three validation tests were performed with meticulous care. According to the results of these validation tests, it was indicated that DF dependence on the aerosol concentration was not because of our experimental system error including measurement instruments but a real phenomenon of the pool scrubbing.

Bubble column is regarded as a kind of wet filtration method. The solution used in filter containment vented system is composed of a large amount of inorganic salt including sodium hydroxide and sodium thiosulfate. Solution with high inorganic salt concentration is more viscous and has stronger surface tension than distilled water. This property has significant effect on bubble size and bubble deformation during formation process. Besides bubble coalescent is suppressed with inorganic concentration increase. This phenomenon is found to possess influence on bubble formation regime when bubble formation regimes belong double, paring and injection regime. The existence of aerosol in solution is another factor that has obvious effect on bubble formation regime and the transition velocity between regimes. This paper apply high speed camera to study the influence rule of inorganic salt and aerosol concentration on bubble size during formation and formation regimes transition.

This reactor uses liquid sodium as coolant owing to its good thermal physical properties, high boiling point and compatibility of cladding material. However, the sodium has a very active chemical properties, for which the free surface of sodium must be protected by inert gas. In the high temperature environment, the sodium atoms diffusion to cover gas slowly, forming a mixed atmosphere that contained large amount of sodium steam. Sodium steam is covered with the free surface of sodium. Then metal sodium will solidify in the inner wall of the pipe or correlative valves with the reduced temperature. This reactor needs to collect and filter sodium steam in order to reduce the hazards to the equipment, piping system, valves and the other devices. Based on the previous research about the purification process of sodium, this paper compared different steam trapping filtration process and carried out the thermal calculation providing basis for research and design of large sodium cooled fast reactor sodium steam trapping filtration process and establishing a reliable sodium steam filtering system.

Even after 6 years since the accident, the exact accident progression for each unit and location of core debris has not been clarified. Currently efforts are directed towards robotic inspection with remote cameras, as well as dose and temperature measurements of the environment inside of the Primary Containment Vessels (PCV). In spite of their effort, the observed environmental data do not support the existence of a large radiation heat sources attributable to the molten core at the bottom of the Primary Containments. Under this situation the author has conducted a forensic engineering study (i.e., different fields of science work together to collect and integrate independent evidences) to clarify the most likely accident scenarios of the Fukushima Daiichi accident. Through this study the author found that the environmental contamination and public exposure could have been substantially mitigated should the following vulnerabilities have been removed before the accident: (1) The threat of hydrogen generation through “radiation-induced electrolysis”. (2) Potential threat of “internal hydrogen explosion” in the suppression pools. The atmosphere on top of the suppression pool water (cover gas) should have been nitrogen. (3) Potential threat of “internal hydrogen explosion” in pipes which had occurred in the case of the Hamaoka Unit 1 accident. (4) The leak rate of PCV should have been testable at its design pressure. Intrinsic safety factor of the containment flanges against effluent leakage should have been rated as a 3 for functional integrity of the PCV. (5) Spread of hydrogen gas from vent lines through duct networks connected to SGTS. The hardened vent line should be independent and provided with filters for Dry Well venting. No rupture disks should have been installed in the vent lines.

The new licensing standards were further improved by taking into account of lessons learned from the Fukushima-Daiichi nuclear accident, and countermeasures against severe accidents were newly required as regulatory items, where severe accidents were defined as serious accidents that occur under conditions exceeding design bases. Organic solvent fire in cell was defined as one of the severe accidents in nuclear fuel reprocessing facilities, which should be investigated, in order to establish methods for evaluating effectiveness of the countermeasures. One of the combustibles in the fire accident at reprocessing facilities is the organic solvent composed of 30% tributyl phosphate (TBP) and 70% dodecane. When the solvent burns, aerosol of soot and radioactive substances are released inside the facility. The aerosol causes a clogging of high-efficiency particulate air filters (HEPA filters) in a ventilation system of the facility, which increases a differential pressure of the filters. We have performed combustion tests simulating the fire accident. As one of interesting results of the tests, we observed, when most of dodecane in the solvent was burned out, a rapid increase in a differential pressure of a HEPA filter, which may cause its rupture. We also found a small amount of RuO 4 release from the burning solvent, which can pass through HEPA filters due to its volatility. These phenomena should be adapted in the effectiveness evaluations of the countermeasures against the fire accident.

Corrosion product in the primary coolant is formed mainly by structure material corrosion and activated corrosion product. Under normal condition, source term of corrosion product maintains at a low level while its radioactive concentration reaches a peak value after the primary is injected with hydrogen peroxide during shutdown oxygenation operation. Following the procedure of corrosion product elimination, analysis has been made on effect which accumulated corrosion product exerts on RCV pre-filter and mixed bed demineralizer. Different primary coolant source terms and the cold shutdown peak value of corrosion product have been considered in the analysis. Under the assumption that dissolved and particle dissolved product has a fixed ratio of 1:1, the surface dose rate of RCV pre-filter varies as the different primary coolant varies. Nuclides proportion has been listed and charted based on the calculation. If cold shutdown condition has been taken in primary coolant source term calculation, then the RCV pre-filter surface dose rate will reach 345Sv/h; If primary coolant design value has been taken in source term calculation, then the RCV pre-filter surface dose rate will reach 42.2Sv/h; If 1/3 primary coolant design value has been taken in source term calculation, then the RCV pre-filter surface dose rate will reach 14.1Sv/h. For the waste resin in RCV mixed bed demineralizer, the corrosion product leaves a small contribution (less than 5%) if the cold shutdown is not considered in primary coolant source term calculation; and a contribution of over 90% if the cold shutdown is considered. The analysis provides a radiation safety analysis basis for NPP solid waste.

Typically, the concentration of natural radioactive aerosols in the containment vessel of nuclear power plant is low and in equilibrium. But when a serious nuclear accident occurred, the massive radioactive aerosols would be rapidly released. In order to ensure the integrality of the containment, the pressure inside the containment must be reduced by reducing the concentration of the aerosol. It can cause a serious damage to the atmospheric environment if such radioactive aerosol directly release. In this paper a new mesoscopic impactor filter has been developed which not only can filtrate and collect the aerosol particles but also can decrease the flowing resistance of gas. This paper intends to make numerical simulation to study the regularity of the deposition of aerosols under the laminar condition at different working condition in mesoscopic impactor filters. The 3D model of the filter was established with the commercial software of ICEM CFD and the meshes were divided accurately. The gas phase uses the laminar model and the particles use DPM (Discrete Phase Model). The detailed modeling method is given and the simulation results are analyzed.

During a severe accident when a corium is ejected from the reactor vessel into the containment, one of the mitigations strategy to contain radioactivity from discharging in to environment and to keep containment under design pressure is to use containment flirted venting system (CFVS). In this paper the key phenomena related to the operation and performance in the components of the CFVS system, pool venture scrubber, cyclone separator, particulate filter, and molecular sieve filters were identified. Based on these phenomena a scaling analysis was performed that was based on system level and local phenomena level scaling. For system level scaling boundary flow, mass flux, pressure and temperature were preserved. On the local phenomena scaling various phenomena were considered. Scaling analysis was also carried out for scrubber system, cyclone and filter. Utilizing a scaled down model from 50 nozzles in the prototype to three nozzles in the scaled model, the flow parameters for the model facility scrubber were obtained. Using these parameters, the governing non-dimensional parameters were obtained for prototype and model facility. The scaling ratios for all the relevant parameters are summarized in this paper.

A Probabilistic Risk Assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan, but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is intended to develop PRA methodology for a combination of low temperature and snow for a Sodium-cooled Fast Reactor (SFR) that uses the ambient air as its ultimate heat sink for decay heat removal under accident conditions. Annual excess probabilities of low temperature and of snow are statistically estimated based on the meteorological records of low temperature, snow depth and daily snowfall depth. To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., plugged intake and/or outtake for the DHRS and for the emergency diesel generator (EDG), unopenable door on the access routes due to accumulated snow, failure of the intake filters due to accumulated snow, possibility of freezing of the water in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around the intake and outtake induced by loss of the access routes, this study has investigated effects of electric heaters installed around the intake and outtake as an additional countermeasure. By using the annual excess probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence is failure of the electric heaters and loss of the access routes for snow removal against the combination hazard at daily snowfall depth of 2 m/day, duration time (snow and low temperature) of 1 day.

An experimental setup has been designed and fabricated for the analysis of filtration performance of the metal fiber filter as applied to Filtered Containment Venting System (FCVS). The main characteristic of this test facility is the presence of the aerosol and Scanning Mobility Particle Sizer. The objective is to investigate the removal performance of the metal fiber filter for aerosol, as well as further understand the filtration process in the metal fiber filter. It is observed that the metal fiber filter is capable of removing more than 99.955% aerosols at the desired flow rate ranging from 0.17 m/s to 0.3 m/s and the resistance has a significant linear correlation with flow rate. Due to the electrostatic effect, diffusion effect, inertia effect, interception effect and gravity effect, most penetrating particle size plays a significant role in removal performance of the metal fiber filter for aerosol. It is also found that with aerosol size ranging from 0.1 μ m to 0.3μm in most penetrating particle size, the filtration efficiency is more than 99.8% at the flow rate of 0.25 m/s. From this study, valuable reference data and useful information are provided for practical applications.

The nuclear U-tube steam generator (UTSG) is a key component for nuclear power plant, and the water level of UTSG must maintain a desired value. In order to overcome the impact of the false water level, an equivalent water level is proposed based on data fusion to simplify the water level control of UTSG. The equivalent water level is composed of steady and dynamic compositions, which are got through the water level sensor of UTSG and flow rate sensors for feed water and steam of UTSG respectively. The steady and dynamic compositions are generated by a low-pass filter and a high-pass filter according to the measuring data characteristics on frequency domain. The false water level characteristics of UTSG have been removed through the equivalent water level application, and then the impact of non-minimum phase characteristic of UTSG has been weakened remarkably. A simulation of water level control is conducted based on the equivalent water level, and the result indicates that a simple PID controller can make the UTSG run with a satisfied water level control performance. It is usually difficult to gain a satisfied control effect through a simple PID controller without equivalent water level, so the measurement method of equivalent water level will benefit the operation of UTSG by simplifying the control process obviously.

The continuous generation of graphite dust particles in the core of a High Temperature Reactor (HTR) is one of the key challenges of safety during the operation. The graphite dust particles emerge from relative movements between the fuel elements or from contact to the graphitic reflector structure and could be contaminated by diffused fission products from the fuel elements. They are distributed from the reactor core to the entire reactor coolant system. In case of a depressurisation accident, a release of the contaminated dust into the confinement is possible. In addition, the contaminated graphite dust can decrease the life cycle of the coolant system due to chemical interactions. On the one hand, the knowledge of the behaviour of graphite dust particles under HTR conditions using helium as the flow medium is a key factor to develop an effective filter system for the discussed issue. On the other hand, it also provides a possibility to access the activity distribution in the reactor. The behaviour can be subdivided into short-term effects like transport, deposition, remobilization and long-term effects like reactions with material surfaces. The Technische Universität Dresden has installed a new high-temperature test facility to study the short-term effects of deposition of graphite dust particles. The flow channel has a length of 5m and a tube diameter of 0.05m. With helium as the flow medium, the temperature can be up to 950 °C in the channel center and 120 °C on the sample surface, the Reynolds number can be varied from 150 up to 1000. The particles get dispersed into the accelerated and heated flow medium in the flow channel. Next, the aerosol is passing a 3 m long adiabatic section to ensure homogenous flow conditions. After passing the flow straightener, it enters the optically accessible measurement path made from quartz glass. In particular, this test facility offers the possibility to analyse the influence of the thermophoretic effect separately. For this, an optionally cooled sample can be placed in the measuring area. The thickness of the particle layer on the sample is estimated with a 3D Laser scanning microscope. The particle concentration above the sample is measured with an aerosol particle sizer (APS). Particle Image Velocimetry (PIV) detects the flow-velocity field and provides data to estimate the shear velocity. In combination with the measured temperature-field, all necessary information for the calculation of the particle deposition and particle relaxation time are available. The measurements are compared to results of theoretical works from the literature. The experimental database is relevant especially for CFD-developers, for model development, and model verification. A wide range of phenomena like particle separation, local agglomeration of particles with a specific particle mass and selective remobilization can be explained in this way. Thus, this work contributes to a realistic analysis of Nuclear Safety.

In the course of a severe accident, a large amount of hydrogen gas is generated by a metal-water reaction in a PCV (Primary Containment Vessel) of Light Water Reactors. Although the filter vent of gas mixture, which includes hydrogen and steam, is an effective method for the accident management of BWR that prevents the PCV overpressure, the filter vent at the early stage of severe accident may cause releasing radioactive material to environment. We have been developing the hydrogen treatment system to prevent excessive pressure without PCV vent and releasing radioactive material to environment. We focus on the oxidation-reduction reaction of metal oxides with high reaction rates, for the hydrogen treatment system. Metal oxide material would be an effective device under low-oxygen conditions like PCV of BWR. The hydrogen treatment system mainly consists of a hydrogen processing unit, a blower and pipes. The hydrogen treatment unit has a lot of reaction pipes in which metal oxides are filled. Some fundamental chemical experiments which we have done have revealed that copper oxides (CuO) rapidly react with hydrogen to form cupper (Cu). Their results show that metal oxides are effective as hydrogen treatment elements. On the other hand, there are few evaluations for the characteristics of hydrogen treatment unit. The dependency of hydrogen treatment performance on gas temperature, hydrogen concentration and pressure is investigated in the present study. We conducted experiments using a test section with one reaction pipe, which simulated a hydrogen processing unit. The processing materials granulated CuO, MnO 2 and Co 3 O 4 with 2mm diameter were used. Gradual increase of processing material temperature in the test section was observed along the gas streams caused by oxidation-reduction reaction after the mixing gases were supplied. Consequently, the hydrogen concentration at the outlet of the test section decreased with time. The increase of the hydrogen reaction rate was also observed with increase of gas temperature, hydrogen concentration and pressure. We have developed the thermal-chemical model of hydrogen processing unit from these experiment results, and confirmed that the model could predict the characteristics of a hydrogen processing unit qualitatively.

Pool scrubbing is one of the effective mechanisms to filter out radioactive aerosols in a severe accident of a nuclear reactor. A lot of work has been done on the pool scrubbing models and experiments. However, large discrepancies still exist between the simulation and experimental results. To improve the pool scrubbing model, an accurate decontamination factor (DF) evaluation by an aerosol measurement and a detailed two-phase flow structure measurement is inevitable. A pool scrubbing experimental apparatus was constructed by the thermohydraulic safety research group in Japan Atomic Energy Agency. The test section is a transparent pipe with the inner diameter of 0.2 m and the length of about 4.5 m. The aerosol laden air flow was injected upwardly into the pool water. The aerosol particle diameter distribution was measured by a light scattering aerosol spectrometer. White polydisperse BaSO 4 particles were used as the aerosol test particles. In the first step, we focused on investigating and reducing the error of DF experimentally. Several problems resulting in the error and their solutions for the error reduction were summarized in this paper. Based on the error reduction methods, the DFs of pool scrubbing were measured in two water submergences. The results showed that the DFs for the aerosol with small diameter were independent of the injecting air velocity in the submergence of 0.3 m. In addition, it was found that the DFs increased with increasing the air flow rate in the submergence of 2.9 m. It was presumed that the increase of DF was dominated by the increase of bubble surface area and/or turbulence intensity with the air flow rate increase, while the effect of the reduced bubble traveling time in the water, which may reduce the DF, was smaller than the increasing effect.

In hazardous industrial activities such as in nuclear facilities, High Efficiency Particulate Air filters (HEPA filters) are essential to ensure the containment of airborne contamination. Most of the filters used in ventilation networks are pleated, in order to offer a larger surface of filtration. For industrial risks likely to lead to an important release of particles (e.g. fire), predicting the evolution of the pressure drop of pleated filters is very important, in order to anticipate any dysfunction, failure or breaking of these devices. Pressure drop variations are linked to airflow rate variations and to clogging process of the medium by airborne particles. Thus, the airflow pattern in a pleat channel is essential for optimizing the filter design and enhancing its lifetime. Particles are transported by the airflow and deposited at the filter surface; hence, the geometry of the dust cake (shape and location) is partially determined knowing the velocity streamlines. The present paper focuses on the characterization of airflows in a clean HEPA filter. The difficulty to perform fine measurement on a real scale filter led us to develop an experimental device, consisting in the reproduction of a single pleat, identical to a real pleat constituting industrial filters. The small dimension of the pleat makes the velocity measurement difficult to establish. That is why μ-PIV method has been adapted to measure the velocity field inside the filter for different filtration velocities at the first moments of the experiment, in order to avoid the impact of clogging by particles used to seed the flow. These particles are DEHS droplets 0.01 < St < 0.05. In the future, these well-characterized airflows will be the basis for CFD computation of particle transport and deposition inside the pleats. Ultimately, the aim is to develop or upgrade physical models predicting the pressure drop evolution of pleated filters, during clogging process in accidental situations.