bluuebunny/crossref_metadata_2025_split huggingface dataset:
Power Distribution in a Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)
Authors: Peiman W. , Pioro I. , Gabriel K. 
Published: February 2014

SuperCritical Water-cooled nuclear Reactor (SCWR) is one of the six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic 1200-MWel pressure-channel SCWR operates at a pressure of 25 MPa with coolant inlet and outlet temperatures of 350°C and 625°C, respectively. High coolant outlet temperature allows for high thermal efficiencies within the range of 45–50%. On the other hand, the high operating temperature of SCWR in turn results in high fuel centerline and sheath temperatures. Hence, it is necessary to determine a power distribution inside a core of a reactor in order to ensure that a fuel and a fuel-bundle design comply with their corresponding temperature limits.
The main objective of this paper is to determine a power distribution inside the core of a generic SCWR by using a lattice code DRAGON and a diffusion code DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermalhydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature of UO2 and UC nuclear fuels and a sheath temperature of a new fuel-bundle design. Results of this analysis showed that the fuel centerline temperature of the UC fuel was significantly lower than that of the UO2. This paper also proposes four energy groups for further neutronic studies related to SCWRs.
Verification of Nonlinear Proper Orthogonal Decomposition Reduced Order Modeling for BWR Fuel Assemblies
Authors: Prill Dennis P. , Class Andreas G. 
Published: February 2014

Thermal-hydraulic coupling between power, flow rate and density, intensified by neutronics feedback are the main drivers of boiling water reactor (BWR) stability behavior. Studying potential power oscillations require focusing on BWR operation at high-power low-flow conditions interacting with unfavorable power distribution. Current design rules assure admissible operation conditions by exclusion regions determined by numerical calculations and analytical methods.
Analyzing an exhaustive parameter space of the non-linear BWR system becomes feasible with methodologies based on reduced order models (ROMs) saving computational cost and improving the physical understanding. A general reduction technique is given by the proper orthogonal decomposition (POD).
Model-specific options and aspects of the POD-ROM-methodology are considered. A first verification is illustrated by means of a chemical tubular reactor (TR) setup. Experimental and analytical results for natural convection in a closed circuit (NCC) [1, 2] serve as a second verification example. This setup shows a strongly non-linear character. The implemented model is validated by means of a linear stability map. Transient behavior of the NCC-POD-ROM can not only reproduce the input data but rather predict different states.
The Research of PHWR Reactor Feeder Flow Measurement
Authors: Sheng Xin 
Published: February 2014

PHWR (also known as CANDU) primary heat transport (PHT) system is to make the pressurized heavy water through the reactor fuel channel, take away the heat produced by the fission of uranium fuel (i.e. fuel cooling), through the reactor feeder pipes, into the Steam Generator. PHWR reactor feeder flow value are a very important parameters, which are stable and meets the requirements of the design, plays an important role in the NPP safe operation.
This paper summarize the Qinshan CANDU reactor feeder flow measurement in practice, based on the feeder flow measurement principle and measurement data are analyzed and studied, including feeder flow measurement stages, site preparation of material and equipment, data measurement methods, data analysis methods, dealing with flow measurement uncertainties, and some other key point need to be attention.
The Qinshan CANDU reactor feeder flow measurement schedules create the same type of CANDU units commissioning records in the world, which is not only shorten the commissioning critical path time, but also laid the foundation for the safety operation. The reasons for the Qinshan success are also to analyze in this article.
Finally, this paper based on the experience of Qinshan CANDU feeder flow measurement, put forward some suggestions on organization and implementation to minimize human error and increase its efficiency, and hope for the help on the similar work area in the future.
A Computational Modeling Tool for Welding Repair of Irradiated Materials
Authors: Feng Zhili , Wolfe Ken , Willis Eric 
Published: February 2014

Weld repairs performed on irradiated reactor components can be problematic due to the presence of entrapped helium in the reactor internal components that is generated by boron and nickel transmutation under neutron irradiation. Under unfavorable conditions of high heat input, high tensile stress and high helium content, sufficient quantities of entrapped helium can migrate to and accumulate at grain boundaries during a welding operation thus resulting in cracks in the weld. Determining crack-free welding conditions solely based on extensive weldability testing for a given welding repair scenario could be very costly, time consuming, and in some cases infeasible.
In this work, a computational model has been developed and sufficiently validated as a cost-effective alternative method for making this determination. This computational model considers the transient temperature profile in the material, the stress history in the material and the helium content. It further includes a methodology for determining the amount of helium that is released under the calculated temperature/stress conditions, a method for calculating the degree to which the helium bubbles grow, and a criterion for determining if the resulting helium bubble distribution will result in cracking as the weld cools. This computational tool is benchmarked with available experimental results on welding of irradiated materials. It is shown that such a model can be a cost-effective tool for determining suitable welding process conditions (heat input and process selection) for repair welding of irradiated materials of reactor internals.
Oxygen Monitoring in the Natural Convection Loop Colonri I
Authors: Di Gabriele Fosca , Kosek Lukas 
Published: February 2014

Oxygen has a fundamental role for the safe operation of GEN IV reactors cooled by Heavy Liquid Metals, HLM. The use of oxygen sensors and dosing the gas in the environment are key issues for the chemistry control of HLMs, in particular when corrosion of structural materials is of concern. In fact, the oxygen concentration must be high enough to grow a protective oxide scale on the surface of the construction materials (steels) in order to prevent their dissolution in the liquid metal. On the other hand, a certain threshold must not be exceeded to prevent precipitation of oxides within the flow paths of the plant. For measuring and controlling the concentration of dissolved oxygen in liquid lead alloys, electrochemical sensors were developed and have been studied for several years.
This study focuses on the work carried out in the CVR, in the convection loop COLONRI I, containing liquid Lead-Bismuth Eutectic, LBE. This vertical loop has several locations where sensors can be placed for monitoring of the local oxygen content. A research study was initiated on the aim of assessing the response and reliability of the sensors in various locations, when different gases were dosed in the liquid metal. For over 1000 hours the sensors were monitored as variables, such as oxygen content and temperature, were changed. Their response as a function of their position was qualitatively evaluated and discussed.
Status of the Third NGNP Graphite Irradiation AGC-3 in the Advanced Test Reactor
Authors: Grover S. Blaine , Petti David A. , Davenport Michael E. 
Published: February 2014

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the NGNP Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment.
The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012 [1]. The third experiment, AGC-3, is scheduled to start its irradiation in late November 2012 and complete in the late summer to fall of 2014. This paper will briefly discuss the design of the AGC-3 experiment and control systems, and present the irradiation results to date.
Centralization of Canada’s Spent Nuclear Fuel
Authors: Nicholson Krista , McDonald John , Draper Shona , Ikeda Brian M. , Pioro Igor 
Published: February 2014

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel.
The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method.
The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.
Coupled Neutronics/Thermal-Hydraulics Analysis of PT–SCWR With 54-Element Bundle Design
Authors: Feng Jia , Shan Jianqiang , Zhang Bo 
Published: February 2014

The radial power distribution is uneven in the Pressure Tube type SCWR bundle because of the higher fuel enrichment and the separation of light water coolant and heavy water moderator. The uneven profile will usually result in high cladding temperature. In order to optimize bundle design, the analysis of coupled neutronics/thermal-hydraulics was carried out.
The neutronics code WIMS-AECL and the subchannel code ATHAS are selected to conduct the coupled neutronics/thermal-hydraulics analysis for the PT-SCWR 54-elements bundle design. The analysis of the reference case showed that the maximum cladding surface temperature of the bundles at both Beginning of the Cycle (BOC) and End of the Cycle (EOC) exceed the design limits. The optimization was carried out through (1) adopting staggered radial uranium enrichment profile and (2) optimizing the pitch-circle diameters of element rings.
The results showed that the optimized bundle can achieve lower maximum cladding surface temperature, which is well below the limiting criteria, and meet the criteria of neutronics demands.
Experimental Investigation of Heat Transfer and Pressure Drop for Turbulent Air Flows in Hexagonal Channels
Authors: Gomez Rodrigo , Krussmann Regina , Böttcher Michael , Arbeiter Frederik , Hering Wolfgang 
Published: February 2014

The heat transfer enhancement through turbulence augmentation is recognized as a key factor for improving the safety and economic conditions in the development of both critical and subcritical innovative advanced gas cooled fast reactors (GFR) and transmutation systems. In the present work, a new experimental facility named L-STAR has been designed and erected at the Karlsruhe Institute of Technology (KIT) to study turbulence flow behavior and its heat transfer enhancement characteristics in gas cooled annular channels under a wide range of conditions. The main objective of the experimental study is to investigate and improve the understanding of complex turbulent convective enhancement mechanisms as well as the friction loss penalties of roughened fuel rod elements compared to smooth ones and to generate an accurate database for further development of physical models. Tests are being conducted in a closed gas loop at various Reynolds numbers with nearly uniform heat release conditions.
The test section consists of an annular hexagonal cross-section channel with an inner electrical heater rod element (smooth and roughened), placed concentrically within the test section, to simulate the flow area of a fuel rod element in a fast gas cooled reactor. In the first step, experimental results of the fluid flow with a smooth heater rod are presented. The pressure drops, as well as axial temperature profiles within the heater rod surface have been measured at Reynolds numbers in the range from 3·103 to 7·104. Experimental program is continued with higher temperatures and the implementation of various artificial surface structures.
Probabilistic Distribution of Parameters Applied to PFM Analysis for Piping in China Experimental Fast Reactor
Authors: Qi Min , Wang Yueying , Liu Jia 
Published: February 2014

The safety assessment method based on probabilistic fracture mechanics (PFM) is well applied to pressure vessel and piping. The PFM analysis is more reasonable and reliable than determinate fracture mechanics (DFM) method. In PFM analysis, the uncertainty of main assessment parameters, such as loads, material character parameters, structure dimension and defect sizes are considered to be random, and the probabilistic distribution of these parameters are determined with the theory of probability statistics.
Related to the practical engineering of China experimental fast reactor (CEFR), this paper has done some research work on the parameters probabilistic distribution, and a method was given to determine the optimum fitting probabilistic distribution function of parameters applied to PFM analysis for piping in the small sample size. The work of this paper makes the foundation of the further probabilistic safety assessment of CEFR piping.
Performance Evaluation of DRACS System for FHTR and Time Assessment of Operation Procedure
Authors: Nakata Junya , Wakui Mikihiro , Mori Michitsugu , Sakashita Hiroto , Forsberg Charles 
Published: February 2014

The Fluoride-salt-cooled High-temperature Reactor (FHR) is a new concept of nuclear power reactor being investigated mainly in U.S. and China. The coolant is a liquid salt with a melting point of about 460°C and a boiling point of over 1400°C. As the baseline decay heat removal system, a passive Direct Reactor Air Cooling System (DRACS) is utilized. Though DRACS system has been developed in Sodium Fast reactors (SFR), there are some differences between both. For example, the system in FHR must decrease heat removal when temperatures are low to avoid freezing of the salt and blocking the flow of liquid. Therefore, considering its characteristics, a numerical investigation of DRACS system is needed to evaluate whether FHR has proper ability to remove decay heat and to be robust for a long-time cooling operation after even a severe accident. Furthermore, in addition to its performance evaluation, it is required to make up the operation plan of FHR considering features of this system. It is highly important, with the view of avoiding severe accident, to determine by when the system should be started up.
In both countries mentioned above, Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is currently in progress to build. Reviewing its design and system is a crucial step needed to develop the FHR technology.
In this research, a performance of DRACS system under some thermal-hydraulic basic events was evaluated by numerical simulation. This paper also suggested the adequate operation procedure suitable for FHTR to avoid a severe accident.
Detecting Changing Balance of Plant Indicators Using Ultrasonic Flowmeters
Authors: Estrada Herb , Phillips Joanna , Hauser Ernie 
Published: February 2014

LEFM (Leading Edge Flow Meter) Ultrasonic Flowmeters have been installed in feedwater pipes in over 68 Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear plants worldwide since the late 1970s. The 8-path, crossed-plane configuration commonly installed in feedwater pipes has been demonstrated through extensive calibrations at nationally certified calibration laboratories to provide feedwater mass flow accuracy better than ±0.3%, regardless of upstream hydraulic conditions. The increased accuracy has been accepted by nuclear regulators in several countries as the basis for an increase in licensed thermal power of nuclear power plants.
LEFM technology provides the ability to methodically check all key contributors to flow measurement uncertainty, thereby providing continuous quality assurance of flow data provided by the system and assuring plant operators that the meters remain within their design basis.
This paper will describe how LEFM chordal ultrasonic flowmeter operation and data quality can be verified, thereby providing the basis for troubleshooting changing balance of plant indicators. The paper will discuss a methodology for estimating changes in moisture content of the steam produced by steam generators at a PWR and will present a recent case study showing how detailed analysis of on-line feedwater flow data has been used to reconcile readings of other plant instrumentation to identify changes in steam moisture of ±0.1%.
Heat Removal Capability of Core-Catcher With Natural Circulation
Authors: Kurita Tomohisa , Tobimatsu Toshimi , Tahara Mika , Aoki Kazuyoshi , Kojima Yoshihiro 
Published: February 2014

Toshiba has developed a core-catcher system. It is to be installed at the bottom of the lower drywell in order to stabilize a molten core flowing down from a reactor vessel. It consists of a round basin made up of inclined cooling channels arranged axisymmetrically, and the structure including risers, downcomers and a water chamber to get natural circulation of the flooding water. So it can cover entire pedestal floor and can work in passive manner.
In order to confirm the heat removal capability of the core catcher with natural circulation, we have conducted full scaled tests in several conditions. Some important dimensionless numbers obtained from fundamental equations of the natural circulation are used for the tests.
Using dimensionless number and to compare with several analysis, we can verify that the experiment is adequate to simulate the actual plant.
The Integrated Aging Management System Applied to Nuclear Power Plants in Korea
Authors: Won Se-Youl , Lee Kyeong-Soo , Lee Jae-Gon 
Published: February 2014

According to Post Fukushima action items in Korea, KHNP has established the integrated aging management system to reinforce aging management of system, structures, and components (SSCs) effectively for seven operating units, which are in service for more than twenty years, and for Kori Unit 1 and Wolsung Unit 1, which are subject to continued operation (CO) based on NUREG-1801 GALL report. KHNP’s integrated aging management programs (AMPs) focus on the establishment of aging management system for long-lived operation of nuclear power plants in Korea.
The integrated aging management system consists of the integrated AMP standard guideline, operating guideline for each plant, individual AMPs of each plant, and AMP Data Base (DB) system including implementation results, basic DB information related to facilities operating in NPPs, and operating information such as operating experience and evaluation report.
The integrated aging management system is importantly utilized for Periodic Safety Review (PSR) and the renewal of CO. Therefore, it will be largely contributed to keep NPPs the level of safety for long time operation through the effective aging management.
Demonstration Test on Concrete With Epoxy Resin Coating Using Ultra-High Pressure Water Jet Decontamination Technology
Authors: Tagawa Akihiro , Tezuka Masashi , Terakura Yoshihiro , Naitou Masayuki , Miyazima Kenji 
Published: February 2014

It is one of the most urgent issues to remediate the nuclear power plants contaminated by radioactive materials discharged following the accident at the TEPCO’s Fukushima Daiichi NPS. Concrete walls of nuclear power plants in Japan are coated an epoxy resin coating for easily performing decontamination. We experimented a cutting test in Fugen Decommissioning Engineering Center using maximum 280MPa pressure and 30L/min water quantity, ultra-high pressure water jet system and 40m3/min air quantity vacuum system. We are conducting a study of decontamination technology in environmental pollution using this decontamination system. This decontamination system has achieved a decontamination factor 10 to 100. In environment experiments, we achieved more than 10 decontamination factor in water pressure of more than 150MPa. Thus, we have confirmed the change in a cutting ability by changing parameters. Parameters are the water pressure, the water quantity and air quantity. The impact force for water jet is a function that contains the test parameters. We have considered it using this function. The results of the test showed that there is a correlation between the impact force for water jet and a cutting capability. This decontamination technology can decontaminate radioactive material of the surface adhesion contamination and reduce the amount of waste generated for a thin cutting.
In addition, we have experimented that the water can be recycled by chemical precipitation. After we experimented flocculation test using aluminum sulfate and zeolite flocculant, we have confirmed that it. can clean water up to the level of suspended solids 5mg/L, in turbid water using zeolite flocculant. This suspended solids concentration can be passed to the water processing system in nuclear power plant. From the test results, we found that ultra-high pressure water jet decontamination technology has a possibility that it can be used for decontamination of Fukushima Daiichi nuclear power plant.