bluuebunny/crossref_metadata_2025 huggingface dataset:
Effects of Prior-Deformation and Water Chemistry on Stress Corrosion Cracking of Austenitic Alloys in High Temperature Water
Authors: Lu Zhanpeng , Shoji Tetsuo , Xia Shuang , Yao Meiyi , Zhou Bangxin
Published: February 2014

The interactive effects of prior-deformation and water chemistry on stress corrosion cracking of austenitic alloys in simulated nuclear power plant coolants were quantitatively investigated. Experimental results showed that increasing material yield strength tends to increase stress corrosion cracking growth rates. Increasing electrode potential tends to increase stress corrosion cracking growth rates of austenitic stainless steels. There is a maximum stress corrosion cracking growth rate for Nickel-base alloys and weld metals at electrode potentials near the Ni-NiO equilibrium line. Crack growth rate of prior-deformed austenitic alloys become less dependent on electrode potential than that of their non-deformed counterparts. The modes of prior-deformation and electrode potential affect the stress corrosion cracking path and growth kinetics. The interactive effects between prior deformation and water chemistry on stress corrosion cracking are analyzed.
Effects of Break Location and Time Uncertainties in Small-Break and Medium-Break LOCA Sequences With Unavailability of HPSI
Authors: Gonzalez-Cadelo J. , Queral C. , Montero-Mayorga J.
Published: February 2014

The Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Regulatory Body, consists of a dynamic methodology of probabilistic safety assessment. It has been applied to a thermal-hydraulic analysis of small-break and medium-break LOCA sequences without HPSI, for different break locations, in a three-loop PWR. ISA methodology allows to obtain the damage domain, defined as the region of the space of uncertain parameters where the damage condition is exceeded, for each sequence of interest. In this work, damage domain relates two uncertain parameters (starting time of secondary-side depressurization and break size) to damage exceedance condition (PCT > 1477 K). Several damage domains have been obtained, each one for each break location of interest (cold leg, hot leg, vessel upper head and vessel lower head). Simulations have been performed with TRACE v5.0 patch 1 code, and the results show the capability and convenience of ISA methodology, in order to obtain accurate results that take into account time uncertainties.
Damage Domains Calculation Associated With Multiple Damages in a Steam Generator Tube Rupture Dynamic Event Tree
Authors: Jimenez Gonzalo , Queral Cesar , Rebollo-Mena Maria Jose , Magan Javier , Fernandez Ivan , Gil Jesus , Sanchez-Perea Miguel , Izquierdo Jose Maria , Melendez Enrique , Hortal Javier
Published: February 2014

The Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermo-hydraulical analysis of a Westinghouse 3-loop PWR plant by means of the dynamic event trees (DET) for Steam Generator Tube Rupture (SGTR) sequences. The ISA methodology allows obtaining the Damage Domains of a SGTR Dynamic Event Tree taking into account the operator actuation times and system response time. The Damage Domains allow to quantify more exactly the risk associated with the sequence even if the contribution to the total risk is due to multiple damages, like dose and core damage in the case of a SGTR. The results shows the impact of the risk associated with the dose as an addition of the risk associated with the core damage. Simulations are performed with SCAIS (Simulation Code system for Integrated Safety Assessment), which includes a dynamic coupling with MAAP thermal hydraulic code.
Selection of Chemical Hazard Prediction Models for Plant Safety and Control Room Habitability Evaluations
Authors: Richmond Mary C. , Wan Ping K. , Dague Brady P. , Davis Kyra W.
Published: February 2014

Assuring the protection of people and the environment from unnecessary exposure to radiation is of great concern to nuclear electric power generators and regulators. In order to secure and maintain a license for operation of a nuclear power plant in the United States, the applicant is required to assess postulated scenarios to determine if an accidental chemical release either onsite or offsite will result in a design-basis event. When determining design-basis events, accident categories such as fire, explosion and/or toxic vapor cloud formation are considered.

Evaluations must consider whether the postulated accidental chemical release will result in operator impairment or damage to safety related structures, systems or components (SSCs), either of which may prevent safe shutdown of the nuclear plant. A critical step in performing such safety evaluations is selection of an appropriate model which will most closely reflect the behavior of a chemical release in the scenario under consideration. Many chemical dispersion models are available for use in safety evaluations for accidental chemical releases; however, it is imperative that the model selected be appropriate for the postulated release conditions. Model selection should be based on careful evaluation of factors such as release location, meteorological conditions, terrain, chemical inventory and storage conditions, as well as the physical and chemical properties of the chemical under consideration in the postulated release.

Failure to select a model suitable for the conditions under which the chemical is released or to appropriately evaluate the physical properties of the chemical and the corresponding limitations of the model may result in underpredicting or overpredicting the impact of a fire, explosion or toxic chemical release. Underpredicting could leave the facility susceptible to damage of safety related SSCs or lead to operator impairment both of which may affect the ability of the plant to safely operate following an accident. While overpredicting the impacts could lead to unnecessary and costly overdesign. This paper will address the considerations that must be evaluated when selecting a chemical dispersion model and will illustrate the importance of model selection in performing nuclear safety evaluations through examples and case studies.
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.
Development and Application of Neutron Damage Calculation Code for ADS System
Authors: Zou Jun , Chen Chong , Zeng Qin , Yang Qi , Chen Zhong
Published: February 2014

A neutron damage calculation code named NDCC was developed for the shielding analysis of ADS. The code can calculate atom displacement, helium and hydrogen production of the nuclides and compositions constituting the structural materials in ADS system. A benchmark was performed in China Lead-Alloy Cooled Research Reactor (CLEAR-I) to test the availability and reliability of the NDCC code. The discrepancy between the NDCC calculation and the results calculated by other codes fell into a reasonable range.
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.
Burnup Computation for HTR-10 Using Layer-to-Layer Movement
Authors: Wu Shang-Chien , Sheu Rong-Jiun , Peir Jinn-Jer , Liang Jenq-Horng
Published: February 2014

This study proposes a layer-to-layer movement model using a once-through fuel cycle strategy to dynamically simulate the on-line refueling process employed in HTR-10. The MCNPX 2.6.0 computer code and continuous energy data library ENDF/B-VII were used in performing all of the computations. In this study, the pebble bed in the core was equally divided into five layers in the axial direction, and the volume fractions of the fuel and graphite pebbles in the initial core were 0.57 and 0.43, respectively. After each fuel cycle, the bottom layer was discharged from the core and discarded while a new layer containing only fuel pebbles was added to the top layer of the core. Hence, the volume fraction of the fuel pebbles increased with greater operation time. This study further proposes that each fuel cycle is stopped to initiate the refueling process for next fuel cycle whenever the effective multiplication factor (keff) reaches approximately 1.005. The results revealed that spikes in the keff versus reactor operation time are the result of burnup and refueling. The fuel cycle tends to approach an equilibrium cycle after refueling five times. In addition, the axial power distribution tends to change from a bottom-peaked to a top-peaked phenomenon as the fuel cycle number increases. In essence, the axial power distribution is nearly un-changed once the reactor core reaches an equilibrium cycle. This phenomenon is also verified by the corresponding axial burnup distribution, average burnup, and mass of special nuclides as a function of operation time.
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.
Experimental Investigation of CCFL in Large Diameter Hot-Leg Geometry
Authors: Al Issa Suleiman , Macián-Juan Rafael
Published: February 2014

CCFL (countercurrent flow limitation) is an important phenomenon for numerous engineering applications and safety of light water reactors. In particular, the possible occurrence of CCFL in the hot-leg of a PWR during SBLOCA or LOCA accidents is of special interest for nuclear safety research. A theoretical review showed that despite numerous experimental works, many scaling and geometrical effects are still not fully understood (channel diameter, inclined riser length, and inclination angle). Since most experimental work has been done in down-scaled hot-leg simulators, it becomes interesting to increase the data base in order to safely extrapolate results to a full-scale hot-leg. Another goal is to provide high quality images of the phase interface for validating CFD codes. There is an increasing interest in performing 3D CFD simulation for CCFL in hot-leg geometries, and thus good experimental data and the development of more representative closure laws for fundamental processes (momentum transfer) are an essential part of the validation and development process. A two-phase flow test facility, COLLIDER, was constructed at the Nuclear Engineering Department at the Technical University Munich in order to investigate air/water CCFL phenomena in PWR hot-leg geometry under atmospheric pressure conditions. The facility concentrates on investigations in large diameter pipe (inner diameter 190 mm) rather than quadratic cross section that although it facilitates optical measurements but does not represent the real geometry. Experimental measurements related to CCFL phenomena are limited in large diameters and hot-leg geometry. COLLIDER represent an approximate 1/3 downscaled model of standard PWR hot-leg geometry with respect to channel diameter, horizontal length to diameter ratio, inclined length to diameter ratio, and 50° inclination angle. First tests were conducted in order to determine the onset of CCFL at different water inlet superficial velocities and for a detailed tracking of the events leading to CCFL occurrence while the gas velocity was gradually increased. Additionally, the deflooding point was determined by gradual decreasing of the gas velocity after CCFL onset in each test run. Consequently, a detailed phenomenological description of flooding/deflooding was obtained besides the important critical gas velocity at CCFL onset and at deflooding in Wallis parameters (JL*0.5, JG*0.5). The results cover low and medium water inlet velocities (JL,in*0.5 = 0.085 → 0.3). Critical gas velocities at CCFL onset show usual trend behavior (decreasing with increased water inlet velocities at low water inlet velocities and increasing with increased water inlet velocities at medium water inlet velocities, see Figure 6). The deflooding line follows a linear tendency quite well. A correlation for the deflooding line based on current results was proposed. Further investigations will include visual observations of the air/water interface for CFD validation.
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.
Effects of the RHR Return Line Elevation to the Suppression Pool Temperature of the Lungmen ABWR Containment
Authors: Chen Yen-Shu , Lin Ansheng , Yuann Yng-Ruey
Published: February 2014

Lungmen Nuclear Power Plant in Taiwan is a twin-unit Advanced Boiling Water Reactor (ABWR) plant. In this study, a long-term GOTHIC model for the Lungmen ABWR primary containment response analysis is established. The wetwell space is vertically divided into several volumes to catch the pool temperature stratification effect. The long-term containment responses for a double-ended feedwater line break (FWLB) accident are calculated. The fuel decay heat is absorbed by the reactor coolant, and the coolant flows to the containment via the broken line. The suppression pool is gradually heated up by the high-temperature gas-water mixture following through horizontal vents. To reduce the pool temperature, the Residual Heat Removal (RHR) system will be required to operate in the suppression pool cooling mode. The RHR pumps have suction flow from suppression pool and discharge it to the RHR heat exchangers for cooling. The cooled water then returns to the pool. An elevated RHR return line is desired to avoid the cooled water being directly sucked again. The wetwell temperature stratification associated with the RHR return line elevation is investigated in this study. Effects of the RHR return line elevation on the pool temperature can be determined since the whole wetwell space is not lumped as a node only. The calculated peak pool temperature is 92.6°C based on the plant piping configuration. The peak temperature can be reduced to 88.9°C by returning the water via the wetwell spray spargers located in the top of the wetwell. However, it should be noted that using the wetwell spray also pressurizes the wetwell because the pool water temperature is higher than that of airspace during the late period of the event. Returning the pool water via the wetwell spray spargers is not suggested because it causes long-term wetwell pressurization.
Thermal Hydraulic Analysis Code Development for DHRS of SFR
Authors: Wang Xiaokun , Zhang Donghui
Published: February 2014

CEFR (China Experimental Fast Reactor) is a typical pool type SFR (Sodium cooled Fast Reactor) which relies on independent positive DHRS (Decay Heat Removal System). This paper describes a new 1-D transient thermal hydraulic code named LR which is based on DHRS of CEFR. LR code can be used for coupling calculations of heat exchange and flow in the primary side of DHX (decay heat exchanger), middle loop of DHRS and air side of AHX (air heat exchanger). The calculations of these three loops are all based on 1-D model for the consideration of the structure of this system, the local heat transfer coefficients and the drag coefficients. Meanwhile, boundary conditions include sodium pool temperature field around the DHX, air temperature field around the AHX and air stack, as well as the condition of the air inlet on AHX. And the initial condition is the steady state which is calculated from the boundary conditions at the initial time. We have simulated the transient process of DHRS in CEFR with this code, the result of which is closed to that of Russian CBTO and fits the conclusions of theoretical analysis.
Temperature Field Analysis for Freezing Sealing Structure of Sodium Gate Valve in China Experiment Fast Reactor (CEFR)
Authors: Hu Lina , Yu Huajin , Lv Mingyu
Published: February 2014

In order to achieve the localization of the large-scale sodium gate valve in fast reactor nuclear power station, this paper use fluid software FLUEN, using two different methods, analysis temperature field of sodium gate valve freezing sealing structure of CEFR. The first method is to establish air model; the second method is to making use of experimental relation to calculate surface coefficient of heat transfer to simulate temperature field of freezing sealing structure of sodium gate valve in CEFR. Compared with measuring results on-site, simulation results of the first method have error within 10%. Therefore, it is feasible that simulating temperature field of freezing sealing structure with this method. That will promote structure design of freezing sealing of sodium gate valve for large fast rector power plant.
Cyclic Softening Effect on Design Margin of JLF-1 Steel
Authors: Li Huailin
Published: February 2014

A reduced-activation ferritic/martensitic (RAF/M) steel, JLF-1, is considered as one of the candidate structure material of the fusion reactors and supercritical water-cooled reactor (SCWR). Low cycle fatigue properties of JLF-1 steel at elevated temperature are the design base to provide adequate design margin against postulated mechanism that could experience during its design life, such as stress range, plastic deformation, and cyclic softening etc. However, the reduction in design margin is significant when the cyclic softening happens in cyclic deformation at RT, 673K, 873K. Thus, for the application as the structural materials, it is necessary to evaluate low cycle fatigue behavior and cyclic softening of JLF-1 steel at elevated temperature since those properties of material at elevated temperature are the key issue for design.
Development of RI-ISI Program for Krško NPP
Authors: Kulat Scott , Graybeal Robin , Montgomery Benjamin , Vrbanić Ivan , Bašić Ivica , Čorak Zrinka , Božič Justin
Published: February 2014

Risk-informed methodologies for inservice inspections of safety related piping in nuclear power plants were formally established in mid-1990s in the U.S. Since then, they have been adopted and applied by almost all of the U.S. plants. Nowadays, risk-informed inservice inspection (RI-ISI) is considered to be a standard for the operating plants in the U.S. It was not long before the RI-ISI practice started to be “exported” from the U.S. to other countries. By now, RI-ISI had found its way to a number of European and other countries. Among the recent examples is the Krško Nuclear Power Plant (Krško NPP), a two-loop Westinghouse-designed PWR located in Slovenia and owned by Slovenian and Croatian utilities. Krško NPP finished its third inservice inspection (ISI) interval in July 2012 and initiated implementation of the RI-ISI program at the start of the fourth interval. The process used to develop the RI-ISI program conformed to the methodology described in Electrical Power Research Institute (EPRI) Topical Report TR-112657 and included a degradation mechanism evaluation, consequence analysis and risk characterization for ASME Class 1 and Class 2 piping, as well as an element/examination selection process, risk impact assessment and inspection implementation program development. This paper describes the development of the Krško NPP RI-ISI program and the results of its RI-ISI application. A discussion is, also, provided on some aspects relevant for application of RI-ISI approaches developed in the U.S to plants outside of the U.S.
Experimental and Prediction Methods Investigation of Two-Phase Frictional Pressure Drop in Rectangular Narrow Channel
Authors: Huang Dong , Gao Pu-zhen
Published: February 2014

An experimental investigation on two-phase frictional pressure drop of water-vapor mixture was conducted in a vertical rectangular narrow channel. The present experimental data were carefully analyzed and compared against the homogenous model, including several well-known two-phase viscosity models. The results reveal that the models using two-phase viscosity proposed by Cicchitti and Dukler provide the relatively accurate predictions, while McAdams, Beattie-Whalley, Lin and Awad-Muzychka poorly predicted the database. Considering the effects of surface tension, gravity, channel size and mass flux, a new correlation for two-phase frictional pressure drop in rectangular narrow channel was developed based on the Cicchitti model. The new correlation predict the experimental data with a mean deviation of 18.7%.
Uncertainty Propagation in Complex Codes Calculations
Authors: Pourgol-Mohammad Mohammad
Published: February 2014

The uncertainty propagation is an important segment of quantitative uncertainty analysis for complex computational codes (e.g., RELAP5 thermal-hydraulics) computations. Different sampling techniques, dependencies between uncertainty sources, and accurate inference on results are among the issues to be considered. The dynamic behavior of the system codes executed in each time step, results in transformation of accumulated errors and uncertainties to next time step. Depending on facility type, availability of data, scenario specification, computing machine and the software used, propagation of uncertainty results in considerably different results. This paper discusses the practical considerations of uncertainty propagation for code computations. The study evaluates the implications of the complexity on propagation of the uncertainties through inputs, sub-models and models. The study weighs different techniques of propagation, their statistics with considering their advantages and limitation at dealing with the problem. The considered methods are response surface, Monte Carlo (including simple, Latin Hypercube, and importance sampling) and boot-strap techniques. As a case study, the paper will discuss uncertainty propagation of the Integrated Methodology on Thermal-Hydraulics Uncertainty Analysis (IMTHUA). The methodology comprehensively covers various aspects of complex code uncertainty assessment for important accident transients. It explicitly examines the TH code structural uncertainties by treating internal sub-model uncertainties and by propagating such model uncertainties along with parameters in the code calculations. The two-step specification of IMTHUA (input phase following with the output updating) makes it special case to make sure that the figure of merit statistical coverage is achieved at the end with target confidence level. Tolerance limit statistics provide confidence a level on the level of coverage depending on the sample size, number of output measures, and one-sided or two-sided type of statistics. This information should be transferred to the second phase in the form of a probability distribution for each of the output measures. The research question is how to use data to develop such distributions from the corresponding tolerance limit statistics. Two approaches of using extreme values method and Bayesian updating are selected to estimate the parametric distribution parameters and compare the coverage in respect to the selected coverage criteria. The analysis is demonstrated on the large break loss of coolant accident for the LOFT test facility.
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.
Improving Thermal-Hydraulic Calculation Modules of THERMIX Code Based on LU Decomposition
Authors: Tang Yang , Zhou Yangping , Zhou Zhiwei
Published: February 2014

THERMIX is a software package for analyzing the thermal-hydraulic and safety behavior of pebble-bed high temperature gas-cooled reactor under both normal condition and accident conditions. The point-wise iterative solution method of THERMIX is time-consuming and difficult to be extended. For the solid-phase thermal conductivity calculation module (THERMIX), gas flow calculation module (KONVEK) and primary loop flow network system calculation module (KISMET) respectively, a global solution method based on LU decomposition with higher efficiency is developed and tested. With good calculation accuracy, the new method has a greater computational efficiency, compared with the original method.
A Study of Phased Array Ultrasonic Velocity Profile Monitor for Flow Rate Measurement
Authors: Fukumoto Takuya , Tsukada Keisuke , Ihara Tomonori , Tsuzuki Nobuyoshi , Kikura Hiroshige
Published: February 2014

A new ultrasonic velocity profile measurement system with a phased array sensor was developed for accurate flow rate measurements. This measurement system employs phased array technique to reduce noise from multiple reflections which occurs in a wedge. However, phased array technique has some difficulties — one of them is about grading lobes. Thus, the array sensor was designed to reduce the grading lobes using simple calculations. Performance of the manufactured array sensor was evaluated by experimental measurements of sound pressure distribution emitted from the array sensor, and optimal parameters to control the array sensor were obtained. Hardware and software of the ultrasonic velocity profile measurement system with a phased array sensor were also developed. The software is based on ultrasonic Doppler method. Velocity profiles of horizontal pipe flow were measured using the developed system, and the availability of the system applying a phased array sensor was confirmed.
The Development of Radiation-Resistant RF Tags for Use at Nuclear Power Plants
Authors: Teraura Nobuyuki , Ito Kunio , Takahashi Naoki , Sakurai Kouichi
Published: February 2014

RF tags based on RFID (Radio-frequency Identification) technology have been widely used in various fields including power plant construction and maintenance for the purpose of improving the identification and traceability of the many components in the facility. To date, various types of tags have been developed, including tags that are resistant to chemicals or high-temperature environments, which are used in specialized fields. When considering widespread use of RF tags in nuclear power plants, there is a concern about the effects of radiation on the RF tags, because the data stored in the tag may receive radiation damage, resulting in corruption of data. Here, we describe a newly designed RF tag that achieves resistance to radiation damage by attaching a radiation shield layer and incorporating automatic data-correction software. This radiation-resistant RF tag has been tested under real radiation exposure fields to verify the intended radiation-resistant functions. It is expected that the use of these radiation-resistant RF tags with a data reader and database system will increase the capabilities of RF tags applied to nuclear power plants and it is also expected to lead to reductions in worker radiation exposure doses.
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.
Analysis of Isolation Condenser Behavior by TRACG Code
Authors: Ishii Yoshihiko , Nakadozono Naoyuki , Ochi Hitoshi
Published: February 2014

For countermeasure of the station black out, an isolation condenser (IC) with enhancing direct-current power supply system is one of effective devices. But not so much transient heat transfer data were obtained for full scale ICs. The IC performance at the accident of the Fukushima daiichi nuclear power plant unit 1 was evaluated by TRACG code. The pressure of reactor pressure vessel was well simulated to measurements.
The Modelling and Coupling Methodology of ANSYS CFX Using Porous Media for PB-AHTR
Authors: Li Linsen , Yuan Haomin , Wang Kan
Published: February 2014

This paper introduces a first-principle steady-state coupling methodology using the Monte Carlo Code RMC and the CFD code CFX which can be used for the analysis of small and medium reactors. The RMC code is used for neutronics calculation while CFX is used for Thermal-Hydraulics (T-H) calculation. A Pebble Bed-Advanced High Temperature Reactor (PB-AHTR) core is modeled using this method. The porous media is used in the CFX model to simulate the pebble bed structure in PB-AHTR. This research concludes that the steady-state coupled calculation using RMC and CFX is feasible and can obtain stable results within a few iterations.