bluuebunny/crossref_metadata_2025 huggingface dataset:
High Temperature Oxidation of Polycrystalline Diamond Coated Zirconium Alloy
Authors: Škarohlíd Jan , Škoda Radek , Kratochvílová Irena
Published: October 2016

Polycrystalline diamond coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys and decrease corrosion rate of zirconium alloy during standard operation. Zirconium alloys are widely used as cladding and construction material in almost all types of nuclear reactors, where usually creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper zirconium alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations (Raman spectroscopy) were done for zirconium alloy covered with polycrystalline diamond layer before and after high temperature steam exposure. Weight increase and hydrogen release ware measured during steam exposure.
The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for In-Vessel Melt Retention Strategies
Authors: Carénini L. , Fichot F.
Published: October 2016

One of the main goals of severe accident management strategies is to mitigate radiological releases to people and environment. To choose the most appropriate strategy, one needs to know the probability of its success taking into account the associated uncertainties. In the field of corium and debris behavior and coolability, research programs are still on going and the possibilities to efficiently cool and retain corium and debris inside the Reactor Pressure Vessel (RPV) then inside the containment are difficult to evaluate. This leads to uncertainties in safety assessments particularly when margins to RPV or containment failure are too weak.

In Vessel Melt Retention (IVMR) strategies for Light Water Reactors (PWR, BWR, VVER) intend to stabilize and retain the core melt in the RPV (as it happened during the TMI-2 accident). This would reduce significantly the threats to the last barrier (the containment) and therefore reduce the risk of release of radioactive elements to the environment.

This type of Severe Accident Management (SAM) strategy has already been incorporated recently in the SAM guidance (SAMG) of several operating medium size Light Water Reactors (reactor below 500MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. A European project coordinated by IRSN and gathering 23 organizations (Utilities, Technical Support Organizations, Nuclear Power Plant vendors, Research Institutes…) has been launched in 2015 with as main objective the evaluation of feasibility of IVMR strategies for Light Water Reactors (PWR, VVER, BWR) of total power around 1000MWe (which represent a significant part of the European Nuclear Power Plants fleet).

This paper intends to show how it is possible to introduce transient evolutions of the stratified corium pool in the evaluation of the heat flux profile along the vessel wall. Indeed, due to chemical reactions in the U–Zr–O–Fe molten pool, separation between non-miscible metallic and oxide phases may occur, modifying the thermal load applied to the RPV. If stabilized stratified corium configurations are well defined and modeled, transient evolutions of material layers in the corium pool are still difficult to predict. The evaluations presented are based on calculations performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. The modeling of transient evolution of corium layers leads to configurations with a thin light metal layer on top of the oxidic one, increasing the so called “focusing effect” (intense heat fluxes on the RPV walls adjacent to the top metal layer). A sensitivity study on some uncertain parameters is proposed to evaluate the impact on the kinetics of layers inversion. Depending on the duration of these transient heat fluxes, the mechanical strength of the RPV could be challenged.
Numerical Simulation of Hydrogen Dispersion Inside a Compartment Using HYDRAGON Code
Authors: Saeed M. , Jiyang Yu , Hou B. X. , Abdalla Aniseh A. A. , Chunhui Zhang
Published: October 2016

During severe accident in the nuclear power plant, a considerable amount of hydrogen can be generated by an active reaction of the fuel-cladding with steam within the pressure vessel which may be released into the containment of nuclear power plant. Hydrogen combustion may occur where there is sufficient oxygen, and the hydrogen release rates exceed 10% of the containment. During hydrogen combustion, detonation force and short term pressure may be produced. The production of these gas species can be detrimental to the structural integrity of the safety systems of the reactor and the containment. In 1979, the Three Mile Island (1979) accident occurred. This accident compelled experts and researchers to focus on the study of distribution of hydrogen inside the containment of nuclear power plant. However after the Fukushima Dai-ichi nuclear power plant accident (2011), the modeling of the gas behavior became important topic for scientists. For the stable and normal operation of the containment, it is essential to understand the behavior of hydrogen inside the containment of nuclear power plant in order to mitigate the occurrence of these types of accidents in the future. For this purpose, it is important to identify how burnable hydrogen clouds are produced in the containment of nuclear power plant. The combustion of hydrogen may occur in different modes based on geometrical complexity and gas composition. Reliable turbulence models must be used in order to obtain an accurate estimation of the concentration distribution as a function of time and other physical phenomena of the gas mixture. In this study, a small scale hydrogen-dispersion case is selected as a benchmark to address turbulence models. The computations are performed using HYDRAGON code developed by Department of Engineering Physics, Tsinghua University, China. HYDRAGON code is a three dimensional thermal-hydraulics analysis code. The purpose of this code is to predict the behavior of hydrogen gas and multiple gas species inside the containment of nuclear power plant during severe accident. This code mainly adopts CFD models and structural correlations used for wall flow resistance instead of using boundary layer at a wall. HYDROGAN code analyzes many processes such as hydrogen diffusion condensation, combustion, gas stratification, evaporation, mixing process. The main purpose of this research is to study the influence of turbulence models to the concentration distribution and to demonstrate the code thermal-hydraulic simulation capability during nuclear power plant accident. The calculated results of various turbulence models have different prediction values in different compartments. The results of k–ε turbulence model are in reasonable agreement as compared to the benchmark experimental data.
The Study on Two Phase Instabilities in a Multi-Channel System
Authors: Lin Yuansheng , Zhang Kelong , Wei Zhiguo , Liu Zhouyang , Yu Rui
Published: October 2016

Commonly two phase flow instabilities always occur in such multi-channel systems as nuclear reactor or steam generator. In this paper, the instabilities occurring in a system with four paralleled channels are investigated under forced flow condition. The system code Relap5 is used to simulate the phenomena in the system. The results show that the system may experience different types of flow oscillation with the increase of heating power uniformly. It is revealed that the thermal hydraulic parameter change dramatically in given channel during the oscillation. Three different heating cases are simulated and studied. The effect of inlet throttling on the stability is also studied and the difference of each channel is analyzed.
Heat Transfer and Fluid Flow Characteristics of One Side Heated Vertical Rectangular Channel Inserting Porous Material
Authors: Shigematsu Takumi , Takeda Tetsuaki , Funatani Shumpei
Published: October 2016

The Very High Temperature Reactor (VHTR) is a next generation nuclear reactor system. The passive cooling system should be designed for the VHTR as the best way of reactor vessel cooling system (VCS). Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR.

Furthermore, we examined the heat transfer of rectangular vertical channel using “Spandrel” panel. “Spandrel” is the metallic plate having grooved patterns. The reason is that we can set it at low cost to the VCS because it is a kind of general construction materials.

The objective of this study is to examine heat transfer characteristics of one side heated vertical rectangular channel with natural circulation and application of spandrel panel to the VCS of VHTR in order to construct the passive cooling system of the VHTR.

We have performed an experiment and a numerical analysis. On experiment, we set the panel to adiabatic wall and supplied 100–400W/m2 heat flux to the panel. In order to obtain the heat transfer and fluid flow characteristics of a vertical channel inserting porous material, we have also carried out a numerical analysis using the commercial CFD code as the first step. From the results obtained in the analysis, it was found that the amount of heat removal was increased for 1–21% by inserting copper wires as porous material.

This paper describes a thermal performance of the one-side heated vertical rectangular channel inserting copper wire with high porosity. From the view point of the economical and safety characteristic, the passive cooling system should be designed for the VTHR as the best way of the system. So, the gas cooling system by natural convection is the one of candidate system.
Numerical Simulation of Fretting Wear of Heat-Transfer Tubes in Steam Generator Under Random Excitation Forces
Authors: Tang Lichen , Qian Hao , Xie Yongcheng , Jing Yi , Huo Yongzhong
Published: October 2016

A numerical simulation method of fretting wear on heat-transfer tube under random excitation is proposed in this paper. Linear springs are used to simulate the in-plane support of anti-vibration Bars (AVBs) to the U-bend tube. The contact forces between the tube and the AVBs are calculated by using FEM random excitation analysis program. And the in-plane relative displacements subjected to turbulence between the tube and the AVBs at each contact points are calculated. The workrate in Archard equation can be calculated by multiplying the contact force, the relative displacement, coefficient of friction and the modal frequency. By using this method, a representative tube is studied. And the effect that the pre-tightening forces between the tube and AVBs had on the fretting wear and the fluid-elastic stability is discussed.
Boiling Heat Transfer Characteristics and Film Boiling Collapse Temperature Through the Two-Dimensional Temperature Field Measurement: Examination of Condition in High Liquid Subcooling Condition
Authors: Takeuchi Hajime , Ohtake Hiroyasu , Ueno Masahiro , Washida Hiroshi , Hasegawa Koji
Published: October 2016

The behavior of rewetting on a high superheated and dry surface, focusing on rewetting temperature just as collapse of saturated and subcooled film boiling was investigated experimentally. Saturated and subcooled impingement jet experiments were conducted by using a silver plate with 40.0 mm × 40.0 mm (length × width) and the distilled water at atmospheric condition. The liquid subcooling was 0, 10, 20, 30, 40, 50, and 75K, respectively. Temperature of the test heater was measured by using a infrared thermometer from a back side of it. Although the liquid velocity and liquid subcooling change, the sputtering temperature, i.e., rewetting initiation temperature, was close to 200 °C predicted by Nishio’s correlation.
The Modification of Fuelling Frequency and the Use of Burnable Absorbers in CANDU Reactors
Authors: Bruce C. L. , Chan P. K. , Song J.
Published: October 2016

CANDU reactors are very novel in that refuelling can occur during normal operation. In fact, on-power refuelling is considered to be CANDU’s primary method of reactivity control. Using a conventional fuelling pattern, refuelling occurs four times per day, and serves to keep certain CANDU reactors functioning within a legislated operating envelope.

As CANDU reactors age, operational margins have begun to narrow. Power ripples that result from the insertion of fresh fuel bundles during on-power refuelling constantly challenge the limit of these boundaries. Therefore, reactivity control devices such as Liquid Zone Controllers (LZCs) are employed to prevent both the violation of these operational parameters and the initiation of a reactor trip.

To address the negative consequences of aging reactors, some research efforts have been directed towards improving CANDU operating margins. Significant gains have been realized from the addition of burnable neutron absorbers (BNAs), specifically 150 mg of Gd2O3 and 35 mg of B2O3 to a CANDU fuel bundle (24 × 106 mg)[3].

This study furthered the work by altering the conventional fuelling scheme in an iterative fashion with the addition of 8 bundle shift rings, in an attempt to reduce the fuelling frequency, and thus the demand on temperamental fuelling machines. Both NU and NU laced with BNAs were considered for this research study.

Results indicated that both the absolute maximum channel power and LZC fill fractions were reduced when 150 mg of Gd2O3 was added to NU fuel; these findings were unanimous for the various fuelling schemes considered, with a fixed fuelling frequency of four times per day. The mean discharge burnup of irradiated fuel containing the BNA also increased slightly for all core following studies. This is reassurance that gadolinium oxide reduces both channel power and LZC levels regardless of the chosen fuelling scheme. Further studies will re-examine the fuelling schemes proposed in this study with a fuelling rate of two times per day with the expectation that gadolinium oxide will successfully facilitate the use of more frequent eight bundle shifts in a CANDU reactor, and reduce fuelling machine usage without imposing a penalty to either burnup or LZC usage.
A Novel Feature Extraction Scheme for NPP Initiating Event Identification
Authors: Lin Ting-Han , Wu Shun-Chi , Chou Hwai-Pwu
Published: October 2016

A novel two-stage feature extraction scheme is proposed in this paper for eliciting discriminant information contained in the data from various nuclear power plant (NPP) sensors to facilitate event identification. Based on the idea of sensor type-wise block projection, the primal features can be extracted without losing the intrinsic structure contained in the multi-sensor data. The features are then subject to further dimensionality reduction by selecting the sensors that are most relevant to the events under consideration. Results from detailed experiments with data generated from a simulator of Taiwan Maanshan NPP illustrate the efficacy of the proposed scheme.
Impact of Testing Activities on Small Modular Reactor Total Capital Investment Cost
Authors: Maronati Giovanni , Petrovic Bojan , Van Wyk Jurie J. , Kelley Matthew H. , White Chelsea C.
Published: October 2016

The use Total Capital Investment Cost (TCIC) as a figure of merit to evaluate the design of a Nuclear Power Plant can help lead to more economically competitive designs. TCIC includes costs of equipment, labor, materials and the associated time value of capital. The team developed the software tool EVAL, which is capable of determining TCIC impacts for any nuclear island (NI) of any design. EVAL was first used to estimate the effect of modularization on TCIC in constructing the Westinghouse Small Modular Reactor (SMR). In particular, three different construction cases were identified. In the first case, modules are manufactured in the fabrication facility and assembled into Super Modules (SMs) in the on-site assembly area, while SMs are assembled in the hole to form the NI. The second case differs from the first case in the fabrication process, as modules are manufactured in the on-site assembly area. In the third case, the NI is ‘stick built’; i.e., the modules are assembled in the hole, where all connections are performed and the structures are built. The analysis highlighted the positive impact of off-site modular construction on TCIC.

EVAL is based on an open evaluation methodology. In this paper, we present an extension of EVAL that aims to analyze the impact of testing on TCIC. As only few Nuclear Power Plants (NPP) were built in the recent years, testing costs and durations are characterized by a high uncertainty. EVAL was used to evaluate the impact of testing on TCIC, considering a range of realistic data points. Testing costs were expressed as a percentage of total labor costs and TCICs were calculated for the three construction strategies.

EVAL was also used to evaluate the impact of modular testing on TCIC. Modularization allows functional testing and system testing activities to be moved from the installation stage to the fabrication and assembly stages, with a subsequent reduction in labor cost and total construction time. TCIC sensitivities were performed on the fraction of testing activities that can be moved from the installation stage. The number of these activities is dependent on both the design and the technologies used during construction. The analysis showed the positive impact of modular testing on TCIC and demonstrated how EVAL can be a tool capable of helping stakeholder decisions.
Experimental Investigation of Spiral Tubes Steam Generator Rupture Scenarios in LIFUS5/Mod2 Facility for ELFR
Authors: Pesetti Alessio , Del Nevo Alessandro , Forgione Nicola
Published: October 2016

In the framework of the EC FP7 LEADER project, an experimental campaign was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the spiral tube Steam Generator (SG) of the European Lead Fast Reactor (ELFR). Two tests are analysed.

The LIFUS5/Mod2 facility implemented a test section composed by 188 tubes, vertically disposed with triangular pitch, in a shell closed by top and bottom flanges and having a perforated cylindrical wall. The central tube injected water at about 180 bar and 270°C, at middle height of the tube bundle, in the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with an argon cover gas at about 2 bar. It was connected to a 2 m3 dump tank, due to the high injection pressure.

In the reaction tank fast instrumentation was set: 6 fast Pressure Transducers (PTs) acquiring data at 10 kHz for precisely characterize the first injection peaks; 70 low constant time Thermocouples (TCs) to understand the vapour evolution path; and 13 strain gages (SGGs) to measure the strain of the bundle and main vessel.

The first test analysed showed a first pressure peak of about 25 bar, due to pressure wave propagation at the cap rupture instant. It did not appear in the second test as consequence of a leakage from the cap before the complete rupture. The following pressurization caused by the entering of water into the reaction vessel was of an analogues magnitude for both the tests (about 30 bar). The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 160°C. The outer rank was cooled down to 340°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests.
Test Section Design for SGTR Experimental Investigation in CIRCE Facility for HLMRS Supported by SIMMER-III Code
Authors: Pesetti Alessio , Tarantino Mariano , Forgione Nicola
Published: October 2016

In the framework of MAXSIMA project, the design of a large-scale Test Section (TS), aiming to experimentally investigate the Steam Generator Tube Rupture (SGTR) postulated event in a relevant configuration for Gen IV MYRRHA reactor, was carried out. The TS will be implemented in the large pool CIRCE facility, at ENEA CR Brasimone. The TS is composed of four tube bundles representing a full scale portion of the Primary Heat eXchanger (PHX) of MYRRHA plant. They allow the execution of four SGTR tests, one at a time, excluding the necessity to extract the TS from the facility after each test. Water is foreseen to be injected at 16 bar and 200°C in the pool, partially filled by LBE at 350°C with a cover gas of argon at about 1 bar.

The pressurization transients of CIRCE vessel and the sizing of the discharge lines and relative rupture disks were numerically predicted by SIMMER-III code on the base of a preliminary simplified configuration of the TS. The obtained results showed that the design pressure of CIRCE main vessel was not reached during more than 10 s of water injection, implementing a singular rupture disk having a diameter of 2 inch activated at 6.5 bar. It appears more than enough to notice, in a real reactor, the occurrence of the SGTR event and stop the water supply, interrupting the accidental scenario. These numerical results were adopted to support the design of the presented TS.
Bayesian Network Technique in Probabilistic Risk Assessment for Multiple Hazards
Authors: Kwag Shinyoung , Gupta Abhinav
Published: October 2016

Conventional probabilistic risk assessment (PRA) methodologies (USNRC, 1983; IAEA, 1992; EPRI, 1994; Ellingwood, 2001) conduct risk assessment for different external hazards by considering each hazard separately and independent of each other. The risk metric for a specific hazard is evaluated by a convolution of the fragility and the hazard curves. The fragility curve for basic event is obtained by using empirical, experimental, and/or numerical simulation data for a particular hazard. Treating each hazard as an independent mutually exclusive event can be inappropriate in some cases as certain hazards are statistically correlated or dependent. Examples of such correlated events include but are not limited to flooding induced fire, seismically induced internal or external flooding, or even seismically induced fire. In the current practice, system level risk and consequence sequences are typically calculated using a Fault Tree Analysis (FTA) that uses logic gates to express the causative relationship between events. Furthermore, the basic events in an FTA are considered as independent. Therefore, conducting a multi-hazard PRA using a Fault Tree is quite impractical. In some cases using an FTA to conduct a multi-hazard PRA can even be inaccurate because an FTA cannot account for uncertainties in events and the use of logic gates limits the consideration of statistical correlations or dependencies between the events. An additional limitation of an FTA based PRA is embedded in its inability to easily accommodate newly observed data and calculation of updated risk or accident scenarios under the newly available information. Finally, FTA is not best suited for addressing beyond design basis vulnerabilities. Therefore, in this paper, we present the results from a study on multi-hazard risk assessment that is conducted using a Bayesian network (BN) with Bayesian inference. The framework can consider general relationships among risks from multiple hazards, allows updating by considering the newly available data/information at any level, and evaluate scenarios for vulnerabilities due to beyond design bases events.
Experimental Investigation of Rewetting During Quenching of Hot Surface by Round Jet Impingement Using Al2O3–Water Nanofluids
Authors: Sharma Avadhesh K. , Modak Mayank , Sahu Santosh K.
Published: October 2016

Impinging jet surface cooling is being used in many industrial and engineering applications due to their higher heat removal rate. Jet impingement is one of the methods to cool hot surfaces, especially in textile, metal and electronic industries. Due to high heat removal rate the jet impingement cooling of the hot surfaces is being used in nuclear industries. During the loss of coolant accidents (LOCA) in nuclear power plant, an emergency core cooling system (ECCS) cool the cluster of clad tubes using consisting of fuel rods. The usual water flow within a reactor core is bottom to top, parallel to the fuel rods. When a hot surface quenched at very high temperature using a jet of cold fluid, during the quenching the initial heat transfer is limited by film boiling. The effective cooling takes place only after the surface temperature is below the leidenfrost temperature.

In the present work an experimental investigation has been carried out to analyze the rewetting phenomenon of a hot vertical stainless steel foil by circular impinging jets of pure water and Al2O3–water nanofluids. The rewetting time and rewetting velocity in the form of dimensionless number (Peclet number) obtained from the thermal images obtained from infrared thermal imaging camera (A655sc, FLIR System). Experiments are performed for different Reynolds number (Re = 5000, 8000), and Al2O3–water nanofluids concentration (Φ = 0.15%, 0.6%)
A Methodology for the Integration of Passive System Reliability With Success Criteria in a Probabilistic Framework for Advanced Reactors
Authors: Brunett Acacia J. , Grabaskas Dave , Bucknor Matthew , Passerini Stefano
Published: October 2016

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies that can be used for the determination of passive system reliability while integrating quantitative success criteria into the risk analysis framework. An updated passive system reliability approach has been developed for utilization in the PRISM PRA that systematically characterizes the impact of passive safety systems on key success criteria. This methodology is derived from the Reliability Method for Passive Systems (RMPS), but is refined to explicitly include consideration of overall mission success through satisfaction of success criteria, rather than only focusing on the passive system itself. This paper provides details on the integrated methodology, focusing on the interface between passive system reliability and success criteria. Specific examples for the passive systems/features of interest, RVACS and inherent reactivity feedback, are included. Additionally, aspects of the integrated passive system and success criteria methodology as they relate to the ASME/ANS Non-LWR PRA Standard are identified and discussed.
Thermal Analysis of the FSP-1 Experiment in the Advanced Test Reactor
Authors: Hawkes Grant L. , Woolstenhulme Nicolas E.
Published: October 2016

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel.

The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Three different types of fuel plates with matching pairs for a total of six plates were analyzed. These three types of plates are: full burn, intermediate power, and thick meat. A thermal analysis has been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A thermal safety evaluation was performed to demonstrate that the FSP-1 irradiation experiment complies with the thermal-hydraulic safety requirements of the ATR Safety Analysis Report (SAR). The ATR SAR requires that minimum safety margins to critical heat flux and flow instability be met in the case of a loss of commercial power with primary coolant pump coast-down to emergency flow. The thermal safety evaluation was performed at 26 MW NE lobe power to encompass the expected range of operating power during a standard cycle. Additional safety evaluations of reactivity insertion events, loss of coolant event, and free convection cooling in the reactor and in the canal are used to determine the response of the experiment to these events and conditions. This paper reports and shows that each safety evaluation complies with each safety requirement of the ATR SAR.
Extraction of {UO2}2+ in Intensified Separators of Different Sizes Using Ionic Liquids
Authors: Tsaoulidis Dimitrios , Garciadiego Ortega Eduardo , Angeli Panagiota
Published: October 2016

In this work the extraction of dioxouranium(VI) from aqueous nitric acid solutions (3M) into TBP/ionic liquid mixtures (30% v/v), is studied in intensified small scale separator channels of different sizes. High speed imaging is used to obtain the important hydrodynamic features of the flow (i.e. plug velocity and plug length) which are related to mass transfer and extraction efficiency. It was found that the overall mass transfer coefficients varied between 0.07 and 0.24 s−1 and increase by decreasing the channel size. The extraction efficiency was also affected by the channel size. A decrease in channel size leads to higher extraction efficiencies, while for a certain channel size the extraction efficiency decreases by increasing the mixture velocity.
A Methodology for the Development of a Reliability Database for an Advanced Reactor Probabilistic Risk Assessment
Authors: Grabaskas Dave , Brunett Acacia J. , Bucknor Matthew
Published: October 2016

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of a reliability database (RDB) methodology to determine applicable reliability data for inclusion in the quantification of the PRA. The RDB method developed during this project seeks to satisfy the requirements of the Data Analysis element of the ASME/ANS Non-LWR PRA standard. The RDB methodology utilizes a relevancy test to examine reliability data and determine whether it is appropriate to include as part of the reliability database for the PRA. The relevancy test compares three component properties to establish the level of similarity to components examined as part of the PRA. These properties include the component function, the component failure modes, and the environment/boundary conditions of the component. The relevancy test is used to gauge the quality of data found in a variety of sources, such as advanced reactor-specific databases, non-advanced reactor nuclear databases, and non-nuclear databases. The RDB also establishes the integration of expert judgment or separate reliability analysis with past reliability data. This paper provides details on the RDB methodology, and includes an example application of the RDB methodology for determining the reliability of the intermediate heat exchanger of a sodium fast reactor. The example explores a variety of reliability data sources, and assesses their applicability for the PRA of interest through the use of the relevancy test.
Simulation of Buoyancy-Driven Flow for Various Power Levels at the NSTF
Authors: Kraus Adam R. , Hu Rui , Lisowski Darius D. , Bucknor Matthew
Published: October 2016

The Reactor Cavity Cooling System (RCCS) is an important passive safety system that is being incorporated in a number of high temperature reactor design concepts. The Natural convection Shutdown heat removal Test Facility (NSTF), located at Argonne National Laboratory, is an experiment with the objective of investigating the flow and thermal behavior of a particular air-cooled RCCS design. It consists of 12 ducts surrounded by a cavity with a heated wall, through which air flows via natural convection before exiting through two chimneys. The NSTF is a ½-scale facility, and is well instrumented in order to provide data for code validation, including Computational Fluid Dynamics (CFD)-grade data in a number of locations. Instrumentation includes fiber-optic Distributed Temperature Sensors (DTS) throughout one of the riser ducts and in the upper plenum. In conjunction with the experimental tests, CFD simulations were performed to support the design and optimization of these natural convection systems. The CFD simulations were performed using the “as-tested” geometry of the NSTF. All CFD simulations were steady-state. Both a full natural convection model and a smaller forced primary flow model were tested. The influence of boundary conditions, notably at the cavity walls, was tested. Initial simulations assumed adiabatic walls but these were later adapted to simulate heat losses, aided by thermal images taken of the exterior NSTF surfaces during testing. Simulations were run for tests at two different power levels. A number of turbulence models were compared to test their influence. Simulation results were compared with experimental data. Convergence was generally good for both models. It was found that the natural convection model was indeed beneficial for correctly estimating local temperatures in a number of areas, particularly near the top of the riser ducts and from DTS measurements along the flow path. Flow in the heated cavity was complex. In general, the experimental trends were predicted well by CFD, although magnitudes could be improved in some areas. The turbulence models tested had a relatively small effect on the shape of the temperature profile in the ducts and on heated surface temperatures. Results from the simulations have been of direct use in improving test procedures and choosing locations for more accurate instrumentation. In future work, full natural convection simulations of more tests will be performed. After this has been completed, best practices can be established for accurately simulating these general types of natural convection systems across a wide range of operating conditions.
A Fluid Film CFD Model for Steam Condensation in Presence of Air
Authors: Bian Haozhi , Sun Zhongning , Ding Ming , Zhang Nan , Yang Yongyong , Tian Wangsheng
Published: October 2016

In a nuclear containment, the mixture of steam and air leads to a significant reduction in wall condensation during Loss of Coolant Accident (LOCA). Detailed studies need to be carried out as the film condensation on the wall is closely rated to the safety of containment. In the present work, a three-dimensional Fluid Film steam condensation calculation method is applied and discussed. The current using steam diffusive based Fluid Film condensation model is introduced particularly and the differences from other condensation models are explained. The validation of Fluid Film model is made by comparing condensation heat transfer coefficient with empirical correlations derived by Uchida et al. (1965). Also, three-dimensional simulations of Su experiments (Su et al. 2013) on steam condensation in presence of air are performed with commercial computational fluid dynamics software which has built in Fluid Film Model. Calculated results of steam temperature and heat transfer coefficient are compared to experimental results (Su, Sun & Zhang 2014) and discussed. The comparison shows that the calculated heat transfer coefficient curve fits well with Uchida empirical correlation and the correlation derived from Su experiment. The deviation from Uchida correlation is within 20% and a maximum deviation of 15% from Su experiment. Detailed field profiles are drawn from three dimensional simulation cases and discussed. It is concluded that the condensation of steam-air mixture gases could be successfully calculated using Fluid Film Model.
Experiment for Wastewater Treatment Technology by Ultrasonic Standing Wave Method
Authors: Hou Suxia , Luo Jijun , He Bin
Published: October 2016

Ultrasonic standing wave, as an innovation separation technologies are used to concentrate the radionuclides and prevent the spread of this hazard to the environment. The principle of ultrasonic standing wave methods for collective separation of radionuclides was discussed, and the model in the ultrasonic field for selective separation of individual radionuclides was researched. In addition, the experimental apparatus for liquid radioactive wastes treatment was designed. In this study, lots of experimental data are used to analysis the influence of the parameters.
CFD Investigation of Wire-Wrapped Fuel Rod Bundle Inner Subchannel Behavior and Dependency on Bundle Size
Authors: Brockmeyer Landon , Carasik Lane , Merzari Elia , Hassan Yassin
Published: October 2016

Sodium fast reactor designs often implement a hexagonal array of fuel rods with wire-wrappers to encourage the exchange of coolant between subchannels. The ability to accurately predict inter-subchannel mixing can be used as a metric for turbulence model performance in capturing wire-wrapped fuel rod bundle flow behavior. In this study inter-subchannel mixing predictions by Large Eddy Simulation (LES) and Reynolds Averaged Navier Stokes (RANS) models are compared. The results indicate that the lower order RANS approach is capable of predicting inter-subchannel mixing inside a 19 rod bundle with acceptable accuracy. The RANS model was extended to 37, 61, and 91 rod bundles to observe the effects of bundle size on inter-subchannel exchange for the center-most subchannels. Transverse velocity magnitude and mass exchange were observed to increase with larger bundle sizes. Inter-subchannel mixing is observed to be a strong function of bundle size for bundles up to 91 rods. The results indicate that the inner subchannels of larger bundles may converge upon a characteristic flow pattern. The 91 rod bundle is not large enough to isolate the inner subchannels from shroud effect, and larger bundles will need to be investigated.
Improvement of Exchanging Method of Neutron Startup Source of High Temperature Engineering Test Reactor
Authors: Sawahata Hiroaki , Shimazaki Yosuke , Ishitsuka Etsuo , Yamazaki Kazunori , Yanagida Yoshinori , Fujiwara Yusuke , Takada Shoji , Shinozaki Masayuki , Hamamoto Shimpei , Tochio Daisuke
Published: October 2016

In the HTTR, 252Cf is loaded in the reactor core as a neutron startup source and changed at the frequency. In this exchange work, there were two technical issues; slightly higher radiation exposure of workers by neutron leakage and reliability of neutron source transportation container in handling.

To reduce the radiation dose by the neutron leakage, detailed numerical evaluations using PHITS code were carried out, the effective shielding method for fuel handling machine was proposed. Easily removable poly-ethylene blocks and particles were used as the neutron shieling, and installed in the cooling paths of the fuel handling machine. As a result, the collective effective dose by neutron was reduced from about 700man-μSv to about 300man-μSv.

As to the neutron source transportation container, the handling performance was improved and the handling work was safely accomplished by downsizing.
Video Monitoring System Availability in Radiation Accident Condition
Authors: Xu Shou-long , Zou Shu-liang
Published: October 2016

Study of availability in radiation environment is presented for video monitoring system. Testing experiment has been completed by seven kinds of Image sensor module which includes three kinds of digital image sensor module and four kinds of analog image sensor module. Radiation accident condition was simulated by γ-ray ionizing radiation environment where the dose rate at 16.63Gy/h 20.20Gy/h and 58.30Gy/h. Availability has been studied by analyzing real time monitoring image quality parameters captured in γ-ray exposure environment. The primary image quality parameters include average gray level of the dark image and synthetic brightness of the color image. The most suitable image sensor module has been selected by image quality parameters for comparison before and after irradiation. Experimental results show that, digital camera has minimum background noise. The radiation resistance of CMOS image sensors is better than CCD image sensors. Therefore, digital video monitoring system with CMOS image sensor has the best image quality parameters and slightest effect in γ-ray ionizing radiation environment where the dose rate is less than 58.30Gy/h. Meanwhile, adequate light could reduce noise interference reduced by γ-ray for all types of video monitoring system. However, digital signal processing integrated circuit board has been destroyed when the accumulated dose reach to 88.40Gy, but the CMOS image sensor integrated circuit board has normal working parameters. As a conclusion, digital video monitoring system with CMOS image sensor could be used for real-time monitoring in radiation accident condition, but the digital signal processing integrated circuit board needs to be hardened by radiation hardened technology.
The Spacer Grid Effect on Heat Transfer at Low Flow Rate in a 5×5 Rod Bundles
Authors: Liu Da , Gu Hanyang , Gong Shengjie
Published: October 2016

It is widely acknowledged that the spacer grid has great effect on heat transfer downstream of it. The conventional correlations to predict the augmentation of the spacer were carried out on high Reynolds numbers. However, recent studies have shown that Reynolds number on the heat transfer enhancement is not negligible when the Reynolds number is lower than about 10000. An experiment to investigate the single-phase convective heat transfer downstream of the spacer grid at low flow rate has been performed in a 5×5 rod bundle. The test section was uniformly heated by a DC power and cooled by water. The Reynolds number covered from about 2000 to 10000. The experiment showed that the existing correlations for heat transfer enhancement by a spacer grid underestimated the maximum enhancement at the grid exit of the spacer grid at low Reynolds numbers. As the Reynolds number decreases, the maximum enhancement increases, nevertheless, when Reynolds number decreases to about 4300, the maximum enhancement tend to converge at a certain value. A new correlation has been proposed to account for the Reynolds number effect on heat transfer enhancement downstream of the spacer grid at low Reynolds numbers and which gave good predictions.
Application of the Analytic Hierarchy Process to Quantitative Evaluation of Nuclear Reactors Decommissioning
Authors: Wan Zhao , Yongling Zhang , Junjie Cao
Published: October 2016

Quantitative evaluation of nuclear reactors decommissioning are difficult because the identification of appropriate quantification methodologies and evaluation procedures is still ongoing. We developed an evaluation framework consisting of 20 indicators derived from the 6 criteria, which could holistically and structurally evaluate nuclear reactors decommissioning. We chose a real decommissioning project in China as a case study. The results demonstrated that the analytic hierarchy process (AHP) was capable of nuclear reactors decommissioning evaluation.
Possibilities of Using Ultrasonic Resonators in the Purification of Nuclear Aerosol Particles
Authors: Hou Suxia , Luo Jijun , Zhang Junjie
Published: October 2016

The present work discussed the possibilities of using ultrasonic resonators to agglomerate radioactive aerosol parties in the air. The mechanism of the nuclear aerosol particles in the ultrasonic chamber was described firstly. Through a careful discussion, the appropriate frequency of the piezoelectric resonator has been selected. Finally, an apparatus based on this was built to verify the effectiveness and applicability of the proposed approach.
Radiation-Induced Electrolytic Corrosion of LWRS: (Part 1) — Basic Mechanism and Implications in Degradation Phenomena
Authors: Saji Genn
Published: October 2016

The author recently found that there should exist a “radiation-induced electrolytic (RIE)” mechanism in the reactor water inducing severe interaction between structural materials and their environments in aged LWRs. This mechanism was identified while trying to theoretically reconstruct the potential differences observed in two in-pile test loops; NRI-Rez in Czech Republic and INCA Loop in Sweden.

These results are indicating that the in-core potential is approximately 0.1/0.4volt higher, in BWR(NWC)/PWR water chemistry respectively, when compared to the out-core regions. Through modeling studies, it was found that the concentrations of (DH)/(DO) for PWR/BWR(NWC) are higher/lower respectively, in the in-core region compared with the out-of-core region. These solute species in high concentrations should spontaneously decompose at the out-of-core region, enabling control of their water chemistry.

This mode of corrosion cell has been dismissed in the nuclear community considering that the transport of ions with flow is insignificant due to high purity of reactor water. Part 1 of this paper focuses on how the RIE phenomena are prompted although the reactor water is kept in high purity. The stable molecular species in the reactor water flow transport the valence electrons. They are released at the cathodic in-core region and are recovered at the anodic out-of-core region. Thus estimated potential differences have been benchmarked with the published in-pile test results for both PWR- and BWR water chemistry environments as explained in Part 2 of this series (1).