bluuebunny/crossref_metadata_2025 huggingface dataset:
Corrosion Inhibitor on Copper and Stainless Steel for Component Cooling Water System of NPP
Authors: Liu Jin-hua , Gong Bin , Jang E , Ma Wei-gang , Wen Ju-hua , He Yan-chun
Published: February 2014

Based on the corrosion issues of component cooling water system (CCWs) in nuclear power plant (NPP), the corrosion inhibition properties and protection mechanism on copper and stainless steel was studied by using tests such as electrochemical method, immersion test and dynamic water simulation. Results show that the optimum inhibitor is the compound of tolyltriazole (TTA) and phosphate, which has an excellent corrosion inhibition efficiency on copper in either pure water or abnormal water. The inhibitor also elevated the pitting potential of stainless steel and contributed to the corrosion resistance.
Measurement of Prompt Neutron Decay Constant of CFBR-II at Near Delayed Criticality Using Randomly Pulsed Neutron Method
Authors: Song Lingli , Li Jiansheng , Zhou Haojun , Jin Yu
Published: February 2014

Prompt neutron decay constant of CFBR-II (China’s Fast Burst Reactor) was measured by the randomly pulsed neutron method when the reactor was at the reactivity of −0.1$. A liquid scintillation detector was used to detect the leakage neutrons and the timing uncertainty of the detection system was less than 3ns. The detector and the Cf-252 fast ionization chamber were placed at several positions. Totally 5 prompt neutron timing distribution curves were obtained and the prompt neutron decay constant was 0.610us−1 in average with the uncertainty of 0.030us−1, which was in good agreement with the M.C. calculation.
Analysis of the Velocity Field of Thermal Stratification in Pipes
Authors: Xing Limiao , Zhuo Wenbin , Chen Bingde , Yuan Dewen , Xiong Wanyu
Published: February 2014

The thermal stratification plays an important role in the aging of the piping because of the excessive stresses caused by the temperature differences in the cross section of the pipe. These stresses can limit the lifetime of the piping, causing deformation, support failure, thermal fatigue, cracking, etc. Several nuclear power plants have experienced such mechanical damages. This phenomenon has been addressed as one of the significant safety issues in nuclear power plant operation. Research of the thermal stratification phenomenon by both Computational Fluid Dynamics (CFD) method and experimental method are performed in this paper. It is observed that in the thermal stratification region, the cold and hot fluid in the pipe flow conversely in the numerical analysis. This phenomenon is caused by the flow characteristic that the surge flow pushes the fluid of the whole cross section downstream, causing pressure drops at the upper or bottom section of the surge flow. The fluid downstream flows backwards because of the pressure difference. In the experiment the converse flow when thermal stratification happens in the pipe is confirmed, which causes a sudden change of the tendency of transient temperature evolutions and the fluctuation of the temperature. Further experimental analysis and validation are needed as this phenomenon enlarges the temperature difference of the pipelines.
Simulation of MHD Effects and Thermal Issues in the DCLL Blanket for ITER
Authors: Chen Long , Du Yue-Fei , Ni Ming-Jiu , Zhang Nian-Mei
Published: February 2014

Coupling thermo-fluid-structure field are analyzed by use of CFD and FEM. A direct simulation of 3D liquid metal flow in the Dual Coolant Lead Lithium (DCLL) blanket is conducted to study the MHD effects and heat transfer influenced by different material properties of FCI (Flow Channel Insert). A consistent and conservative scheme and PISO method on an unstructured collocated mesh are employed to solve the incompressible Navier-Stokes equations with the Lorentz force included based on an electrical Potential formula. The finite element method is employed to study mechanical behavior of FCI. The velocity distribution, MHD pressure drop, current stream lines and temperature distribution of blanket, thermal deformations and thermal stresses in FCI under external magnetic field are investigated. The effects of thickness and electric conductivity of a silicon carbide FCI on temperature field in fluid-structure coupling field, thermal stresses and deformation are analyzed. This work is the theoretical basis of optimizing blanket.
Modeling TBM Components by Using RELAP5 Code
Authors: Zhang Yaoli , Peng Tianji
Published: February 2014

RELAP5 is a best estimate system code suitable for the analysis of all transients and postulated accidents in Light Water Reactor systems. It was usually used to solve plant thermal-hydraulic problems on system scale. RELAP5 was used to study thermal-hydraulic behavior on component level in this paper. The Test Blanket Module (TBM) was a key component in Chinese Helium-Cooled Solid Breeder (CN HCSB) system. One sub-module of TBM was simulated by RELAP5/MOD3.4. The flow paths, Be pebbles neutron multiplier as well as Li4SiO4 pebbles tritium breeder of TBM were modeled by using hydrodynamic components models as well as heat structure models provided by RELAP5. Steady-state condition was studied and the results were compared with CFD results provided by Fluent code. The steady-state results were in consistent with CFD results when the sub-module was well modeled by RELAP5. The results showed that RELAP5 could be used to solve thermal-hydraulic problems on component scale when the component was well modeled. With a detail-modeled TBM, the transient conditions of CN HCSB system could be simulated more precisely by RELAP5.
A Neutronics Concept Design of Lead-Bismuth Cooled Accelerator-Driven System for Minor Actinide Transmutation
Authors: Li Xunzhao , Wu Hongchun , Cao Liangzhi , Zheng Youqi
Published: February 2014

Pursuing a high minor actinide (MA) transmutation rate, this paper proposes a neutronics concept design of lead-bismuth (LBE) cooled accelerator-driven system (ADS) with burnup reactivity swing less than 1% and proton beam current smaller than 17mA. After a comparison with other types of fuels, Uranium-free metallic dispersion fuel (TRU-10Zr)-Zr* is selected to obtain a harder neutron spectrum to transmute MA. With a MA initial loading, the suitable proportion of initial Plutonium to transuranium element (TRU) is found around 33% to make sure that the burnup reactivity swing is less than 1%. The location of the spallation target is optimized to guarantee high external spallation neutron source efficiency and to lower proton beam current. For the subcritical system, initial effective multiplication factor is 0.97, and the thermal power is 1000 MW. For the accelerator, proton with energy of 1.5GeV and a parabolic spatial profile is provided by proton linac. It is demonstrated by the numerical results that the transmutation rate of MA is about 28% after 600 effective full power days (EFPD) while the support ratio for LWR units with the same power is about 46.
3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code
Authors: Gosmain Cécile-Aline , Rollet Sylvain , Schmitt Damien
Published: February 2014

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel.

EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source).

The code needs as input:

- for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI;

- the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology.

This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.
Camera Elevator Control Design for Manipulator Crane in PWR
Authors: Zhang Lei , Zhang Peng
Published: February 2014

Camera elevator is the dominating device of the manipulator crane closed circuit TV system which is used in under constructing domestic PWR in China. The manipulator operator can use camera elevator to view refueling operation (insertion and withdrawal) and other underwater condition. This document expatiates on the independent design of camera elevator control system which is integrated on manipulate crane control system. The success of the design will provide conveniences to the operator to let him to operate both hoist and camera elevator. To contribute to the invention of manipulator crane control system in PWR, the control of camera elevator, which is the main part of the system, is developed totally. The camera elevator control system is mainly composed of power distribution part, PLC control part, HMI control part, Servo-motor drive part, position measuring part, weight measuring part and safety protection part. The control design includes requirement corresponding, electrical product selection and configuration, PLC software program development, Servo-motor drive software setting, HMI and control panel design and so on. The document is very detailed to describe the control design process and emphasizes the software development. The logic controller is consisted of two systems, one is standard logic controller and other one is safety controller. The PLC software is the most important technique which includes whole control logic of device and all interlock protect function. The Servo-motor drive software can guarantee motor has good dynamic characteristic and short response time. The HMI design can provide most of operational information, fault alarm display, interlock and bypass operations etc. The control design has been validated by lots of test to assure the camera elevator’s function is up to related specification requirement finally.
The Comparative Analysis of Hermoluminescence Personal Dosemeter Distribution Method in Group Plant Mode
Authors: Zhou Tienan , Hu Ling , Huang Hong
Published: February 2014

The paper focuses on the distribution of the Hermoluminescence Personal Dosemeter in the nuclear power plants which have been in operation, and analyzes the strengths and weaknesses of the existed methods in group plant mode. By summarizing the operation and management experience of the existed distribution methods, It puts forward a way to reduce the investment of dose monitoring equipments and the investment of operation and maintenance input. This new method is called centralized distribution. But this method also has some inadequacies, the paper puts forward some specific suggestions (including management methods and techniques) to consummate the new method, so that the centralized distribution method is more effective and feasible.
Electromagnetic Self-Locking Device for Air Cylinders in Spent Fuel Storage System of Pebble-Bed High Temperature Gas-Cooled Reactor
Authors: Wu Bin , Wang Jinhua , Li Yue , Liu Jiguo
Published: February 2014

In the spent fuel storage system of pebble-bed high temperature gas-cooled reactor, several air cylinders would be employed in complex machines, such as the spent fuel charging apparatus and the spent fuel canister crane. The cylinders were designed to actuate movements smoothly in radioactive environment. In order to lock them in safe position when the compressed air source is offline by accident, an electromagnetic self-locking device was designed. When power-off, the compressive spring would push out the lock plunger to enable self-lock. When power-on, the lock plunger would be withdrawn by the magnetic force of the coil to unlock the cylinder. In order to optimize the design more efficiently, numerical simulation was performed to optimize geometry parameters of the structure surrounding the working air gap so as to improve the performance of the device. A prototype was then fabricated. Combining the simulation results with experimental test, the actuating force characteristics of the device in locking and unlocking process was analyzed. The temperature rise when the device stays unlocked with power supply was also calculated and validated. The results showed that this electromagnetic self-locking device could realize the locking and unlocking functions effectively, and the maximum temperature rise also conforms the required limit. The as-fabricated device would help guarantee the fail-safe feature of the air cylinders of complex machines in compressed air outage.
Magnetic Field Effect on the Natural Convection of a Liquid Metal in a Horizontal Cylinder
Authors: Wang Z. H. , Yang S. , Chen H.
Published: February 2014

Magneto-hydrodynamic (MHD) effects are widely exploited in different industrial processes. MHD play an essential role in nuclear fusion, where it is involved in the behavior of the liquid metal alloys employed in some of the currently considered designs of tritium breeding blankets. Results of numerical simulations are presented for the natural convection of a liquid metal placed in a horizontal cylinder in the presence of a vertical magnetic field. When there is an additional magnetic field, an induced current is produced by the movement of the liquid metal in a magnetic field. Induced current and magnetic field interaction produced a Lorentz force which is opposite to the movement direction and inhibits the natural convection and heat transfer intensity. The numerical results show that the magnetic field has a observable effect at the heat transfer process of the liquid metal natural convection in a horizontal cylinder. An interesting effect of the magnetic field during the internal flow is the deceleration of the flow velocity for liquid Gallium. The magnetic field in the horizontal radial direction, which is perpendicular to the natural convection caused by the temperature gradient all the while, has the most significant influence on the natural convection, while the influence on the axial direction is comparatively weak in medium magnetic field. With the increase of the intensity of magnetic field, the inhibition is more obvious.
AP1000 3D Rod Ejection Reactor Core Thermal Analyses
Authors: Qin Huimin
Published: February 2014

Due to more and more studies were made on fuel and clad failure mode in international nuclear industry, some acceptance criteria of rod ejection accident should be revised in order to consider the fuel and cladding behavior change caused by radiation. Furthermore, along with the rapid development of computer hardware and improvement in physics analysis methodology, it is necessary and it is capable to use three-dimensional method to perform rod ejection accident analysis in order to relax some conservatism. In this paper, 3D method has been applied to AP1000 equilibrium core load pattern to perform fuel temperature and fuel enthalpy analysis.
Nuclear Power Plant Design Verification Based on the Analyzer
Authors: Wang Hanlin , Zheng Mingguang , Qi Tunfeng
Published: February 2014

In this paper, the current nuclear power plant (abbreviated as NPP) design verification methods are analyzed, and employing nuclear power plant design analyzer (abbreviated as NPPDA,) is proposed to improve design verification. The main points include: the basic concepts of NPPDA, a detailed discussion of the principles of how NPPDA could play a role in design verification, and extended considerations on application of NPPDA in NPP design verification.
Assessment of Choking Flow Models in RELAP5 for Flashing Flow Through Small Cracks
Authors: Revankar Shripad T. , Wolf Brian , Riznic Jovica R. , Srinivasan Ganesh
Published: February 2014

The estimation of leak rates through steam generator tube crack is an important safety parameter. An assessment of the choking flow models in thermal-hydraulics code RELAP5 is performed and its applicability to predict choking flow rates through steam generator tube cracks is addressed. A RELAP5 nodalization was created to model experimental data from literature. It is found that both the Henry-Fauske and Ransom-Trapp models better predict choking mass flux for longer channels. As the length of a channel decreases the both models’ predictions diverge from each other. While RELAP5 has been shown to predict choking flow in large scale geometries, it is not suited well for small channel lengths. In the case of a more conservative approach, where over prediction of mass flux through short channels is best, the Henry-Fauske model would be most appropriate.
Evaluation of Damaging Effects on a NPP Integrity Subjected to Beyond Design Earthquake
Authors: Lo Frano Rosa , Forasassi Giuseppe
Published: February 2014

The aim of this study is to evaluate the influence of the damaging or failure effects induced by seismic loadings on the loading bearing capability of the overall nuclear structures, systems and components (SSCs). In particular it was analyzed the behaviour of the SSCs in terms of their required safety functions and capacity assuming the occurrence of an earthquake having magnitude beyond the design basis value (in agreement with the “stress tests” suggestions, presently foreseen by the European and International Associations).

To the purpose, quite refined FEM models were set up and implemented considering suitable materials behaviour and constitutive laws for the reactor materials, particularly for the concrete material behaviour (which could suffer failure and damage mechanisms during the ground shaking). Several type of damaging mechanics were considered.

The obtained results seemed to confirm the overall containment reliability even if with minor upgrading actions in operating procedures and design characteristics. Moreover they allowed to appropriately check the NPP containment strength reserve.
Experimental Study of Feed Water Level Decreasing Effect on VVER Steam Generator Model Operation in Condensation Mode
Authors: Morozov A. V. , Remizov O. V. , Soshkina A. S.
Published: February 2014

The essential technological distinctions of the AES-2006 project with VVER-1200 reactor, equipped with passive safety systems, determine some specifics in the character of accidents with the coolant leaks from the reactor primary circuit and the refusal of active part of the emergency core cooling system.

In the case of an accident with depressurization of the reactor primary circuit, the system of passive heat removal (PHRS) ensures the transition of steam generators (SG) into the mode of steam condensation. As a result, a condensate comes to the core, providing its additional cooling. But in the case of beyond design basis accident the rupture of second circuit pipelines or PHRS steam-condensation path is possible. At such type of accidents the feed water level in SG vessel will be decrease, what can results in the deterioration of its condensation power.

For experimental investigation of the condensation mode of VVER steam generator, a large scale HA2M-SG test facility was constructed in IPPE. The test facility incorporates VVER reactor SG model with volumetric-power scale of piping is 1:46, PHRS heat exchanger imitator, cooling by process water and buffer tank, equipped by steam supply system from the IPPE heat power plant. The facility main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project.

Experiments at the HA2M-SG test facility have been performed at the pressure 0.36 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. On the base of the results of these experiments the correlation of condensation power from feed water level in SG model was obtained. The results of carried out tests make it possible to draw a conclusion about sufficiently large stability of VVER steam generator working in condensation mode to feed water level decreasing.
Experimental Observation of the General Trends of the Steady State and Instability Behavior of Supercritical Water Natural Circulation
Authors: Lv Fa , Huang Yan-ping , Wang Yan-lin , Yan Xiao
Published: February 2014

Experimental study on steady state and instability behavior of subcritical and supercritical water natural circulation was performed in a rectangular loop (NPIC-SCNCL) in NPIC, with the pressure of 10 and 21.1∼24.8MPa. According to the experimental results, in steady state the natural circulation displayed higher mass flow rate, heating power and outlet water temperature of the heated zone under supercritical pressure (22.2∼24.8Mpa) than those under subcritical pressure (10MPa). The effects of heating power and water pressure on the steady state behavior of supercritical water natural circulation were also carried out. In the experiments, natural circulation showed instability both in supercritical and subcritical conditions. The results showed that the natural circulation instability appeared when the outlet water temperature of the heated zone got close to the critical temperature under supercritcial pressure. In comparion, the natural circulation instability appeared when the outlet water temperature of the heated zone got close to the saturated temperature in subcritical condition. The instability behavior of supercritical water natural circulation showed the fluctuations of the mass flow rate, water temperature and pressure et al. In the experiments, we also observed that the fluctuations of the parameters mentioned above for supercritical water natural circulation instability could disappear when the heating power is big enough which seemed that the natural circulation returned to a new steady state.
Scientific Bases of Water Chemistry for Corrosion Control of NPPs by Integration of Radiation- and Electro-Chemistry
Authors: Saji Genn
Published: February 2014

In this paper, the author continues his investigation on the scientific basis of water chemistry specifications by applying his recent theory, which integrates the elemental radiation- and electro-chemistry reactions in the “Butlar-Volmer equation.” The B-V equation is well established as the basic material balance equation in corrosion science.

The author’s new approach has been compared with the published in-pile test results of the electrochemical potential differences between the in-flux and out-flux regions for both the PWR- and BWR water chemistry environment. Although the theoretical estimation generally reproduced the experimental results, there remains significant deviation from the experimental results at the very low DH region (<10cc-STP/kg-water) in PWRs as well as the low DO region (<10ppb) in BWRs. Although these regions are outside of the water chemistry specifications of general interest, the scientific causes of the deviation must be clarified.

In this paper, the author found that the deviations are due to the dominant radiation-chemical reactions involving hydrogen ions and hydrogen peroxide at the lower ends. Although the radiation- and electrochemical reaction was further exploited with respect to the potential differences induced by the hydrogen peroxide, the effects were disappointingly small, when estimated in terms of a mixed potential of the electrode reactions. This leads the author to suspect that hydrogen-ion-radical reactions should be the main causes. Currently further analyses are in progress.
A Post Accident Safety Analysis Report of the Fukushima Accident — Future Direction of Evacuation: Lessons Learned
Authors: Saji Genn
Published: February 2014

In nuclear safety, the source term is introduced to provide adequate isolation of the nuclear hazards from the public, by establishing a concept of ‘effective distance.’ This combines a geographical distance to the site boundary and an effective ‘distance’ with the use of engineered safety features (i.e., a containment system and its cooling system), combined with evacuation procedures to prevent radiation injury. Severe accidents occur when these safety systems failed to function. This basic safety approach was once again jeopardized by the Fukushima accident, which followed the Chernobyl accident. The factors that mitigated the effluent releases however depend greatly on the intrinsic safety features combined with the accident management. The multi-layered retention/decontamination factors that a nuclear power plant possesses should be incorporated in specifying the environmental source term. The Fukushima accident provides a reasonable upper bound with respect to environmental releases due to a LUHS (loss of ultimate heat sink), which triggered a prolonged SBO (station blackout).

Due to the anticipated radiological consequences, the Japanese Government issued a series of evacuation orders, resulting in the evacuation of approximately 160,000 people from the Fukushima area. The prolonged evacuation is believed to be the cause of over one thousand “disaster-related (pre-mature) deaths (DRDs)” which have been reported among the evacuees due to psychosomatic effects (48%) and the disruption of medical and social welfare facilities (18%). In the future these types of deaths should be avoided.
Preliminary Design and Analysis of a Lead-Bismuth Natural Circulation Loop
Authors: Ran Guangming , Huang Qunying , Gao Sheng
Published: February 2014

Heavy liquid metals (HLMs) such as pure lead and its alloys are being studied worldwide as potential candidate coolants for future advanced nuclear systems due to their good neutronics and thermal-physical properties. Owing to the large thermal expansion coefficient, HLM-cooled nuclear systems have a significant natural circulation capability which is desired in reactor designs for improving the passive safety of reactor systems. Some advanced HLM-cooled reactor designs even apply natural circulation for primary cooling as normal operation mode. Moreover, natural circulation could be greatly enhanced by injecting an inert gas (e.g. Argon gas) into the hot leg of the coolant circuit. Although HLM-cooled reactors based on natural circulation or gas-lift mechanism have attractive advantages compared with traditional reactors utilizing mechanical pumps, some great challenges and uncertainties remain due to the lack of operation experiences. However, very few experiments have been carried out so far for HLM natural circulation or gas-lift studies. More efforts need to be carried out, especially on flow regime characterization, heat transfer coefficient and flow stability in natural circulation. For the purpose to study the basic mechanisms of HLM natural circulation as well as gas-lift pump, design activities of a lead-bismuth natural circulation loop were carried out in FDS team. In the present study, preliminary design and analysis of the loop were carried out. According to the present design, the maximum natural circulation flow rate in the loop was foreseen to reach 1.38kg/s with a limiting input power 30kW. The analysis results of natural circulation, gas-lift driven circulation as well as gas-injection enhanced circulation in the loop showed a similar trend to meet the power law with exponents differing from the buoyant mechanisms.
A Numerical Simulation on Effects of Distance Between Bubble and the Wall on Behavior of Rising Bubble
Authors: Huang Ying , Gao Puzhen
Published: February 2014

A numerical investigation of two-dimensional air bubble behaviors under the effect of gravity in still water based on the VOF (Volume-Of-Fluid) method is carried out.

Initially, the surface tension effects on the behavior of the bubble is analyzed, which contains the simulation of the ascending motion of a single air bubble in liquid and the study of the interaction between bubbles in terms of coalescence.

Additionally, the differences of single bubble’s rising motion in an infinite surroundings and in a vertical narrow channel are analyzed. The coalescence of bubbles is also studied. The motion of bubbles with different diameters in a vertical channel is simulated. It is found that the bubbles’ behavior depends on the distance between the bubble and the wall.

Finally, numerical simulation of the motion of several bubbles of the same size, at the same initial horizontal position and with uniform distribution is carried out. The result reveals that the bubbles at different distances from the wall have different velocities, after a while, the bubbles distribution presents as “U”.
Model Passivated Carbon Electrodes for Fluorine Generation in MSRs and the Nuclear Fuel Cycle
Authors: Seto Kelvin S.-H. , Ikeda Brian M.
Published: February 2014

Elemental fluorine, F2, is used in the nuclear fuel cycle for the isotopic separation of uranium-235 and 238, as well as for the purification of LiF-BeF2 in molten salt reactors. F2 is generated on an industrial scale by an electrochemical process using carbon electrodes in a KF-2HF molten salt. Carbon electrodes are used for industrial F2 generation due to its chemical stability, high conductivity, and relatively low cost. One of the main issues faced when using carbon electrodes in this chemical system is passivation through the formation of C-F compounds on the surface of the electrode. This results in a loss of anode wettability to the electrolyte and diminished charge transfer rate. The voltage needed for the fluorine evolution reaction increases which negatively impacts the safety of the system, increases the operating costs, and leads to faster degradation of the electrode.

The degradation of electrical properties during passivation is progressive, eventually leading to electrode deactivation. The process of deactivation begins with a passivating C-F layer at potentials above the equilibrium potential (2.92 V). The layer is both non-wetting to the KF-2HF media and insulating. Deactivation begins with inhibited F2 bubble detachment, formation of a persistent gas layer, and finally deactivation as the electrode surface is completely covered by a thick, insulating C-F layer causing charge transfer to cease. Only a small current is able to flow, even at high potentials (up to 9 V), indicating F2 generation is completely inhibited. The purpose of this study is to manufacture and test model carbon electrodes and, to examine the non-wetting properties of a partially fluorinated surface. The electrodes will be prepared by mixing PTFE-particles with Vulcan carbon powder and then pressing to form pellets. These electrodes should have a reproducible surface for electrochemical performance studies that will lead to a better understanding of the surface chemistry. The research will develop novel electrodes with a goal to minimize the voltage required for F2 production. This will enhance the efficiency in the overall process and lower the manufacturing costs for F2.

Carbon electrodes with different PTFE-content (20 w.% and 35 w.%) were synthesized. Electrochemical fluorination was then carried out at different potentials in the F2 generation region (4 to 8 V) in molten KF·2HF electrolyte at ∼90 °C. The electrochemical behaviour of the carbon-PTFE electrodes was examined and compared for both fluorine passivated and non-passivated graphite, amorphous carbon, and vitreous carbon electrodes. The growth of the electrical double-layer capacitance between the carbon electrodes and the KF·2HF molten salt was studied. The effects of composition of fluorinated and non-fluorinated carbon on electrode performance are presented.
Natural Uranium Equivalent (NUE) Fuel: Full Core Use of Recycled Uranium and Depleted Uranium in CANDU Reactors
Authors: Cottrell Catherine M. , Boubcher Mustapha , Chen Mingjun , Zhang Zhenhua , Kuran Sermet
Published: February 2014

Since 2008 Candu Energy and the CANDU® utility in China (CNNC TQNPC) partnered to convert the Qinshan CANDU reactors to use recycled and depleted uranium based fuels, a first-of-its-kind. A fuel composed of both components and simulating natural uranium behaviour was developed, designed, manufactured and verification tested in the Qinshan units with success. This paper discusses the full core implementation and licensing work that will culminate with the conversion of Qinshan CANDU units to Natural Uranium Equivalent (NUE) fuel use by early 2014.
Safety Performance of a Large Scale Sodium Cooled Fast Reactor in Unprotected Transient Overpower Scenarios
Authors: Wu Yuanyu , Yu Hong , Ren Lixia , Hu Wenjun
Published: February 2014

Design options of large scale sodium cooled fast reactors are being studied intensively in China, and China Fast Reactor 1000 (CFR-1000) is one of the options. An unprotected transient overpower (UTOP) scenario analysis is carried out during the concept design phase of CFR-1000, aiming at estimating the safety performances of the core during the transient and providing feedbacks for the design. Reactivity insertion scenarios are simulated with different reactivity insertion rates. Safety aspects of core design are evaluated based on the simulation and analyses, including the design of reactivity feedbacks, the design of control rod as well as safety margin of fuel and cladding. The UTOP scenario analysis provides insights of CFR-1000 core design in safety aspect, and valuable information on safety design is given for future China fast reactor design.
Verification of Control Rod Assembly Homogenization for LMFBR With a New Lattice Physics Code GALAXY-H
Authors: Kamiyama Yohei , Koike Hiroki , Kirimura Kazuki , Yamaji Kazuya , Kosaka Shinya , Matsumoto Hideki
Published: February 2014

A new FBR lattice physics code GALAXY-H has been developed by Mitsubishi Heavy Industries, Ltd. (MHI). GALAXY-H is a hexagonal version of GALAXY, which is a two dimensional transport calculation code for PWR assembly. GALAXY-H generates assembly nuclear constants used in the FBR core calculation code. The methodology of flux calculation for GALAXY-H is based on the method of characteristics (MOC) as well as GALAXY. The fuel assemblies of Japanese demonstrated and commercial FBRs are intended to contain the inner duct called FAIDUS where molten fuel is removed to prevent re-critical at severe accident. One of the objectives for developing GALAXY-H is to treat the inner duct and wrapper tube configurations exactly. In this paper, the method generating nuclear constants of control rod assembly is developed with multi-assembly model to exclude the super-cell model that has been used in FBR design so far. Besides, GALAXY-H employs the SPH method for reduction of homogenization error, which is popular method in LWR design. From this, the advanced nuclear constants calculation method for FBR control rod assembly is developed and the basic applicability of FBR nuclear design by using GALAXY-H is confirmed.
Validation of the System Code APROS for Fast Transients
Authors: Raqué Manuel , Herbell Heiko , Schulenberg Thomas
Published: February 2014

The thermal-hydraulic system code APROS Version 5.09 [1] is being applied in the European project SCWR-FQT to evaluate the performance of the safety systems for a nuclear test facility operated with supercritical water. In order to validate the commercial code for predictions of transient phenomena, two adequate hydraulic experiments from literature have been simulated.

The experiment of Fujii and Akagawa [2] investigated hydraulic shocks as they will occur in case a pipe, which is stationary passed through by water, is abruptly closed. A simple numerical model with adapted time and space nodalization was able to reproduce the observed physical phenomena, such as the magnitude of the initial pressure wave and reflection time, in detail.

In a second experiment, a test series was performed by Becker et al. [3] and Mathisen [4] in order to examine the natural circulation in a closed loop for different heating rates and system pressures. For a stepwise power increase, the typical mass flow characteristic for boiling channels was recorded until the onset of flow oscillations. In further runs, the effect of different initial conditions on the flow stability was analyzed. This paper compares the numerical predictions with both experimental results. The numerical models could describe the physical phenomena with appropriate accuracy.
Shaping Public Opinion on Nuclear Power
Authors: Xu Shenjun , Xu Jing , Sun Jing , Qiu Li , Wang Ru , Zhao Bin
Published: February 2014

After the Fukushima Daiichi nuclear disaster, the public acceptance of nuclear power has dwindled to historical low. Governments were forced to cancel and postpone new projects or even shut down reactors in operation due to an increased anti-nuclear sentiment. This paper aims to provide an international perspective of how various factors can affect public opinion of nuclear power.

In this paper, we rebut the previous-held argument that nuclear education is conducive to the public support of nuclear power. It is found that the relationship between educational efforts and public support is captured by a downward-sloping line. The paper then assesses the effect on the public acceptance of demographics, socioeconomic status, political environment and risk orientation using correlation coefficients table. The largest public concern comes from the insecurity of nuclear power plants and radioactive materials. The health of an economy also plays a major role in determining people’s attitude toward building new nuclear power plants.

The paper also suggests some solutions for each category of countries based on the research analysis.
Investigation of Characteristic-Temperature Approaches for Heat-Transfer Correlations at Supercritical Conditions
Authors: Mokry Sarah , Pioro Igor
Published: February 2014

It is expected that the next generation of water-cooled nuclear reactors will operate at supercritical pressures (∼25 MPa) and high coolant temperatures (350–625°C). In support of the development of SuperCritical Water-cooled Reactors (SCWRs), research is currently being conducted for heat-transfer at supercritical conditions. Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare-tube data can be used as a conservative approach.

A number of empirical generalized correlations, based on experimentally obtained datasets, have been proposed to calculate Heat Transfer Coefficients (HTCs) in forced convective heat transfer for various fluids, including water, at supercritical pressures. These bare-tube-based correlations are available in various literature sources. There have been a number of methods applied to correlate heat transfer data. The most conventional approach, which accounts for property variations in the data, is to modify the classical Dittus-Boelter equation for forced convection. However, analysis and comparison of these correlations has shown that differences in HTC values can be up to several hundred percent.

In general, the familiar correlations of Dittus-Boelter and Bishop et al. have used the bulk-fluid temperature approach for characteristic temperature properties evaluations. However, at high heat fluxes, fluid near the tube-wall will have a temperature close to that of the wall temperature. This might be significantly different from the bulk-fluid temperature. Therefore, another approach can be used based on the wall temperature as the characteristic temperature. The Swenson et al. correlation is based upon this approach. Finally, a third approach has been considered in which the film-temperature is used as the characteristic temperature (Tf = (Tw+Tb) / 2). McAdams et al. based their correlation for annuli on this approach.

Therefore, the objective of this paper is to evaluate the three characteristic temperature approaches, (1) Bulk-fluid temperature approach; (2) Wall-temperature approach; and (3) Film-temperature approach, and determine which characteristic temperature method can most accurately predict supercritical water heat transfer coefficients. Both classical correlations and more recently developed correlations are considered in this investigation.
CFD Analysis for the Dome Section of Maanshan Nuclear Power Plant
Authors: Cheng Wen-Chi , Ferng Yuh-Ming , Chen Shin-Ru , Chieng Ching-Chang
Published: February 2014

This study aims to reveal the possibility of a crack due to the temperature gradient on RPV head in Maanshan Nuclear Power Plant. The temperature distribution inside RPV and on the RPV wall is calculated by Computational Fluid Dynamics (CFD) software STAR-CCM+. The computational domain starts from the bottom part of the upper plenum and continues throughout the entire dome region, so that the high temperature coolant leaking to the dome region through gaps between solids can be included to calculate the leak flow and temperature distribution in the dome region. In the present study, Computational cells are generated for half of the upper plenum region and dome region using symmetry assumption to save computer resources. The total number of cells is ∼65 millions including 3 prism layers. The calculations not only provide the detailed information of flow and temperature distribution inside the dome region but also the stress distribution on the dome surface. A small variation of ∼0.8Pa of stress distribution is obtained by the adiabatic assumption of all solid structures.
Development of SUBSPACE-Based Hybrid Monte Carlo-Deterministic Algorithms for Reactor Physics Calculations
Authors: Zhang Qiong , Abdel-Khalik Hany
Published: February 2014

This paper presents an innovative hybrid Monte-Carlo-Deterministic method denoted by the SUBSPACE method designed for improving the efficiency of hybrid methods for reactor analysis applications. The SUBSPACE method achieves its high computational efficiency by taking advantage of the existing correlations between desired responses. Recently, significant gains in computational efficiency have been demonstrated using this method for source driven problems. Within this work the mathematical theory behind the SUBSPACE method is introduced and the performance is demonstrated based on a fixed-source problem. Furthermore, the SUBSPACE method has been successfully extended to address core wide level k-eigenvalue problems. The method’s efficiency is demonstrated based on a three-dimensional quarter-core problem, where responses are sought on the pin cell level. The SUBSPACE method is compared to the FW-CADIS method and is found to be more efficient for the utilized test problem because of the reason that the FW-CADIS (Forward Weighted Consistent Adjoint Driven Importance Sampling) method solves a forward eigenvalue problem and an adjoint fixed-source problem while the SUBSPACE method only solves an adjoint fixed-source problem. Based on the favorable results obtained here, we are confident that the applicability of Monte Carlo for large scale reactor analysis could be realized in the near future.
More on Design Evaluation of Nuclear Power Piping Subjected to Time-History Dynamic Loads According to ASME B&PV III
Authors: Zeng Lingfu , Jansson Lennart G.
Published: February 2014

A nuclear piping system which is found to be disqualified, i.e. overstressed, in design evaluation in accordance with ASME B&PV code, Section III, can still be qualified if further non-linear design requirements can be satisfied in refined non-linear analyses in which material plasticity and other non-linear conditions are taken into account. Our work presented earlier in ICONE16-20 categorized the design rules in the code into linear and non-linear design rules and attempted to clarify the corresponding design requirements. In this paper, an in-depth review of these rules and relevant requirements for time-history dynamic loads designated in Service Limit Levels C and D, such as those caused by e.g. water-hammer effects, is given. In particular, several changes of these rules in the current edition of the codes and their applications are addressed, and several rules which are intensively applied but often found to be inconsistent and confusing in Sweden are discussed. It is suggested that the design evaluation can be made directly through non-linear transient finite element analysis using a strain-based criterion for time-history dynamic loads of either reversing or non-reversing type.
Introduction to KEPIC (Korea Electric Power Industry Code)
Authors: Sohn Myoung-Sung , Joo Hyun-Jae , Byun Hoon-Seok , Kim Jong-Hae
Published: February 2014

Korea Electric Power Industry Code (KEPIC) has been developed on the basis of referring to the prevailing U.S. codes and standards which had been applied to the electric power facilities in Korea.

KEPIC has been applied to the construction of Ulchin Nuclear Units 5&6 since 1997 as per the endorsement of Ministry of Science and Technology, and is being completely applied to the construction of Shin-Kori and Shin-Wolsung Nuclear Power Plants (NPPs) in Korea and of Barakah NPPs in UAE.

Although the technical requirements of KEPIC related to design and inspections for pressure vessels have been developed on the basis of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Korea Electric Association (KEA) that is KEPIC maintenance organization is trying to find and improve unreasonable requirements based on the acquired experiences for the continuous construction and operation of NPPs.

In the administrative requirements, KEPIC, unlike ASME, adopted not only mechanical field but also electrical field related to NPPs. Furthermore, the qualifying system for services such as Nondestructive Examination, Heat Treatment, Design and Equipment Qualification Organization was adopted to improve safety and reliability for NPPs.

In the technical and administrative requirements, KEA has been performing betterment studies such as ‘integration of welding coordinator in welding quality system of KEPIC’, ‘induction of Korean national standards (KS) materials equivalent to ASME/ASTM materials into KEPIC-MH codes’. The result of study for development of Steel-plate Concrete (SC) structure and modularization technique for NPPs has been adopted to 2010 edition of KEPIC-SNG.