bluuebunny/crossref_metadata_2025_split huggingface dataset:
In-Plant 3D Positioning System Using Point Cloud Data for Remote Decontamination Machine
Authors: Satoh Yoshinori , Aikawa Tetsuro , Kawaguchi Yuji , Sakamoto Naoya , Saito Masahiro , Kaneda Masayuki , Sakai Hitoshi 
Published: October 2016

Toshiba Corporation, a member of International Research Institute of Nuclear Decommissioning (IRID), has contributed to decontamination works throughout Fukushima Daiichi Nuclear Power Plant from outside ground to inside of the buildings. Speedy decontamination works allow workers to access and stay inside of the contaminated buildings for many hours, and as a result, all decommission works can be accelerated. Some remote decontamination machines to decontaminate the inside of the reactor buildings from a remote safe building have been developed for workers not to be radiated by high-level radiation. Conventionally, operators have remotely controlled the decontamination machine through multiple views of some remote surveillance cameras mounted on it. Because the position data such as GPS data is not available in the buildings, it was hard for operators to detect its absolute position and orientation in the building, and it took much time to recognize targets to be decontaminated. In order to reduce positioning time and make operation works easier, we constructed 3D positioning system to automatically detect the absolute 3D position of the decontamination machine in the reactor buildings from a remote safe building. Moreover, we can also keep records of decontamination works easily by tracking 3D position of the decontamination machine.
This paper shows the overview of our approach of 3D positioning and a result of examinations in the mock-up facility that simulates a part of the inside of the reactor building at Fukushima Daiichi NPP.
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.
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.
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.
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.
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%)
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.
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.
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).
Experimental Characterization of Airflow Within a Clean HEPA Filter Used for the Containment of Radioactive Contamination in Nuclear Facilities
Authors: Youssef Alilou , Soleiman Bourrous , Dominique Thomas , Nathalie Bardin-Monnier , Philippe Nérisson , Thomas Gélain 
Published: October 2016

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.