bluuebunny/crossref_metadata_2025 huggingface dataset:
Large Eddy Simulations of Jet Flow Interaction Within Staggered Rod Bundles
Authors: Salpeter Nathaniel , Hassan Yassin
Published: April 2011

The present work investigates the turbulent jet flow mixing of downward impinging jets within a staggered rod bundle based on previous experimental work. The two inlet jets had Reynold’s numbers of 11,160 and 6,250 and were chosen to coincide with available data [Amini and Hassan 2009]. Steady state simulations were initially carried out on a semi-structured polyhedral mesh of roughly 13.2 million cells following a sensitivity study over six different discretized meshes. Very large eddy simulations were carried out over the most refined mesh and continuous 1D wavelet transforms were used to analyze the dominant instabilities and how they propagate through the system in an effort to provide some insight into potential problems relating to structural vibrations due to turbulent instabilities. The presence of strong standing horseshoe vorticies near the base of each cylinder adjacent to an inlet jet was noted and is of potential importance in the abrasion wear of the graphite support columns of the VHTR if sufficient wear particles are present in the gas flow.
Management of Single Point Vulnerability Components
Authors: Yu Yijun , Xu Ligen , Wang Xin , Wang Liansheng
Published: April 2011

This paper indicates the importance of classified management of components in view of different functions of plant components, presents the principles based on which Third Qinshan Nuclear Power Plant (TQNPP) implements the classified management of components, and introduces the concept of SPV component in nuclear power stations. It focus on expounding the analysis and identification of SPV systems and components, explaining the methods to determine the list of SPV key systems and to evaluate the system priority sequence; getting the SPV fault tree of the system and SPV points; classification for the SPV points, finding the SPV component chain. The management requirements and practice for SPV components are discussed from different viewpoints of component management.
Design and Simulation of Solenoid for Buncher of a Linear Accelerator
Authors: Ebrahimi Basabi Ehsan , Davani Feryidoon Abbasi
Published: April 2011

One of the most important components of a linac is buncher. Throughout most of the buncher, the electrons are well forward of the crest and have velocities considerably less than light velocity, thus they are in a region of radial defocusing and a considerable fraction of the beam will be lost unless defocusing action is counteracted by some other applied forces. The simplest way to do this is to set up a longitudinal magnetic field which interacts with the radial motion of the electrons and causes them to follow helical orbits through the space occupied by the field. In this paper, five solenoids were designed to provide necessary magnetic field inside the buncher. Magnetic field was analytically calculated and compared with simulation results of CST. Because of resistance in wires, some amount of energy appears in form of heat, so heat power was calculated analytically and cooling system was designed for these solenoids by ANSYS.
Neutron Radiography System Collimator Design via Monte Carlo Calculation
Authors: Jafari Hamid , Shahriari Majid
Published: April 2011

Neutron radiography uses the unique interaction probabilities of neutrons to create images of materials. This imaging technique is non-destructive. MCNP Monte Carlo Code has been used to design an optimized neutron radiography system that utilizes 241Am-Be neutron source. Many different arrangements have been simulated to obtain a neutron flux with higher amplitude and more uniform distribution in the collimator outlet, next to image plane. In the final arrangement the specifications of neutron filter, Gamma-ray shield and beam collimator has been determined. Simulations has been Carried out for a 5Ci 241Am-Be neutron source. In this case 43.8 n/cm2s thermal neutron flux has been achieved at a distance of 35cm from neutron source.
Modeling the Spectral History in the Depletion of a PWR Core
Authors: Bilodid I.
Published: April 2011

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS.
Synthesis of Results Obtained With Laser Cutting, a Promising Dismantling Tool
Authors: Pilot Guy
Published: April 2011

Several experiments have been carried out with a Nd-YAG laser as a future dismantling cutting tool. At the beginning, industrial lasers could cut only small thicknesses due their small power (1 kW) but now, with the rise in industrial power (8 kW), it is possible to cut plates up to 100 mm thickness. This technique is practical with the use of optical fibers which allow to maintain the laser generator in a non-radioactive zone. This paper provides a synthesis of the measurements of aerosols and gases produced in different configurations (cutting in air and underwater) and some comparisons with other cutting tools.
Quantification of Fault Tree Models for Initiating Events
Authors: Wakefield Donald J. , Epstein Steven A.
Published: April 2011

The computation of initiating event frequencies has received much attention in recent years. This attention reflects the importance of support system failures in the computation of core damage frequencies from internal events for nuclear power plants. Multi-train support system failures, such as a total loss of service water or a total loss of component cooling water, potentially represent significant impacts for many nuclear plant designs. The contribution to basic event importance from such initiator models has also been recognized, especially as part of the Mitigating Systems Performance Index (MSPI) program; Reference 1.
Diversity-Oriented FPGA-Based NPP I&C Systems: Safety Assessment, Development, Implementation
Authors: Kharchenko Vyacheslav S. , Bakhmach Eugenii S. , Siora Alexandr A. , Sklyar Vladimir V. , Tokarev Viktor I.
Published: April 2011

Key challenges caused by implementation of diversity-oriented approach and FPGA technology are discussed in context of NPP I&C systems safety. National and international standards containing the requirements to diversity application in NPP I&C systems are analyzed. A few evolution stages of multi-version NPP I&C systems (Reactor Trip Systems) are described taking into account different types of version redundancy (hardware, software, FPGA diversity). Main attention is attended to the methods of increasing tolerance of NPP I&C systems to physical and design faults using multiversion technologies. A life cycle model and multi-version technologies of FPGA-based I&C systems development are analyzed. Implementation results of safety-critical NPP I&Cs developed by RPC “Radiy” using FPGA technology are described. The FPGA-based platform RADIY™ ensures scalability system functions, dependability and diversity. More than 20 different FPGA-based I&C systems were successfully developed, produced and implemented on the NPPs of Ukraine and Bulgaria during last five years.
Current Status and Development of Core Calculation Methodology for Sodium Cooled Fast Reactor in Mitsubishi
Authors: Ikeda Kazumi , Moriwaki Hiroyuki , Nakazato Wataru
Published: April 2011

This paper presents current status and development of nuclear calculation methodology for sodium cooled fast reactor in Mitsubishi and two core design codes; an original two dimensional cell/lattice calculation code, PIJHEX and a thermal hydraulic calculation code, MIX-MKII. Mitsubishi Atomic Power Industries, Inc. developed several core design codes for Monju and JOYO and recently this activity continues still further for the next FBR in Japan. It is explained in this paper that the PIJHEX is confirmed to be valid from the comparison of Monte Carlo code and the examination on physical phenomenon in calculation results. Besides, the approach toward design and development of code system in Mitsubishi and MIX-MKII are introduced briefly.
Fault Detection and Identification in NPP Instruments Using Kernel Principal Component Analysis
Authors: Ma Jianping , Jiang Jin
Published: April 2011

In this paper, kernel principal component analysis (KPCA) is studied for fault detection and identification in the instruments of nuclear power plants. We propose to use mean values of the sensor reconstruction errors of a KPCA model for fault isolation and identification. They provide useful information about the directions and magnitudes of detected faults, which are usually not available from other fault isolation techniques. The performance of the method is demonstrated by applications to real NPP measurements.
An Overview for Achieving Public Understanding and Acceptance of Nuclear Power: Bangladesh Perspective
Authors: Mollah A. S.
Published: April 2011

Nuclear power is a safe, clean and economic energy source. The growth of the nuclear power option is impeded in many countries by public concerns over the safety and environmental consequences of producing electricity by means of nuclear reactors. Nuclear power is more compatible with the environment through reduction in emission of green-house gases, fuel diversification, and energy security. Public concern has been expressed in most countries about the construction and operation of nuclear power plants, and this public concern has in many cases led to postponement or failure to start or expand nuclear power programs, and in some cases even caused a retrenchment of existing programs. This paper examines the nature and causes of public concerns about the development nuclear power and the need for public understanding and acceptance of nuclear energy. Some preliminary results on public opinion survey on nuclear energy in Bangladesh are presented in this report. Preliminary survey shows that, Bangladeshi people have a quite satisfactory rate of support to nuclear energy development, which exceeds 60%.
The Constructability Analysis of AP1000 Containment Vessel Bottom Head
Authors: Gao Nianci , Fang Xiaopeng
Published: April 2011

Nuclear power is a proven, safe, plentiful and clean source of power generation, and AP1000 is the only Generation III+ reactor to receive Design Certification from the U.S. Nuclear Regulatory Commission (NRC). One of the major features of this advanced reactor is the utilization of the passive safety systems and a steel Containment Vessel (CV). As the boundary isolated to keep the radioactive waste to release to outside and also the medium to transfer the heat, CV is a mainly part of the Passive Containment Cooling (PCS) System, which is the key to move the resident heat from the core to the air. The CV is made up with Sub-Assemblies, such as Bottom Head (CVBH), 1st Ring, 2nd Ring, 3rd Ring, and Top Head (CVTH), which are assembled with steel plates. So to build the Containment Vessel on site, we have to assemble each individual part with steel plates fabricated in shop first, transport the Sub-Assembly to Nuclear Island (NI) and lift them in place. Right now, the first AP1000 in the world is under construction at Sanmen, China. Based on the statement above, the construction of CVBH of the 1st AP1000 is a notable achievement and implementation without experience proven, the constructability analysis about work sequence, detailed logic, activity duration, production rate and improvement of them during the process is worthy to study.
Reduction of Dynamic Response of Buildings, Equipment and Pipe Networks Using SERB-SITON Method to Encrease NPP Safety Margins
Authors: Serban Viorel , Panait Adrian , Androne Marian , Ciocan George Alexandru , Zamfir Madalina
Published: April 2011

Buildings, equipment and pipe networks, herein below called “structures”, within nuclear and classic objectives are affected by the dynamic actions of earthquake, shocks and vibrations type, herein below called“ excitations”. The repeated action of excitations on structures most often lead to important built-up of kinetic and potential energy in oscillating systems made-up of such structures. Such a built-up is leading to tens of times increase of the amplitude of the dynamic response in accelerations and displacements of structures as to the excitation amplitude. In its turn, such an increase is leading to the exceeding of the efforts and distortions of the structures material, accompanied by the occurrence of damages or destroy. Because in most cases the excitation cannot be reduced or eliminated, the only technical solution available for the engineers is to reduce the dynamic response of the structures on the existing excitations. The innovative SERB-SITON solution developed and applied within the Subsidiary of Technology and Engineering for Nuclear Projects (SITON) is based on the use of SERB-SITON isolation behavior devices with low friction and controlled stiffness on horizontal plane to cut-off the dynamic action transfer from the excitation to the structure and telescopic devices with controlled elasticity and damping both to reduce the dynamic response of the structures, as well as the dissipation of the energy transferred to the structure or the limitation of structure distortion upon the dynamic actions. The paper presents the performances of SERB-SITON mechanical devices used to protect buildings, equipment and pipe networks against dynamic actions and an adequate method to evaluate the dynamic response of structures (by using SERB-SITON devices), in real time inclusively. For large-size buildings such as nuclear power plants, old buildings, churches, bridges, etc., SERB-SITON isolation devices with friction by sliding or rolling with a very small and adjustable, stiffness installed under the structure on any horizontal direction, are presented. For reinforced concrete or metal framework buildings such as high buildings, industrial halls, towers, etc., SERB-SITON telescopic devices with controlled stiffness and damping large, force inclusively are presented. For equipment and pipe networks, SERB-SITON supports that are capable to overtake large permanent loads with relative displacements on two directions for thermal displacements, and also capable to elastically overtake and damp dynamic actions, are presented.
Thermal Aspects of Using Uranium Nitride in SuperCritical Water-Cooled Nuclear Reactors
Authors: Grande Lisa , Peiman Wargha , Mikhael Sally , Villamere Bryan , Rodriguez-Prado Adrianexy , Allison Leyland , Pioro Igor
Published: April 2011

SuperCritical Water-cooled nuclear Reactors (SCWRs) utilize a light-water coolant pressurized to 25 MPa with a channel inlet temperature of 350°C and outlet temperature of 625°C. Previous studies have indicated that uranium dioxide (UO2) nuclear fuel may not be suitable for SCWR use, because the maximum fuel centerline temperature might exceed the industry accepted limit of 1850°C. This research paper explores the use of uranium nitride (UN) as an alternative fuel option to UO2 at SuperCritical Water (SCW) conditions. A generic 1200-MWel Pressure-Tube (PT) -type reactor cooled with SCW was used for this thermalhydraulics analysis. The selected fuel option must have a fuel centerline temperature not higher than the industry accepted limit of 1850°C. Furthermore, the sheath (clad) temperature must not exceed the design limit of 850°C. The sheath and bundle geometry were adopted from previous studies. A single fuel channel was modeled using the UN fuel and an Inconel-600 sheath for several Axial Heat Flux Profiles (AHFPs). Uniform, upstream-skewed cosine, cosine and downstream-skewed cosine AHFPs were used. For each AHFP bulk-fluid, sheath and fuel centerline temperatures, and Heat Transfer Coefficient (HTC) profiles were calculated along the heated length of the channel. The calculations show that the UN fuel maintains a centerline temperature well below the industry accepted limit due to its high thermal conductivity at high temperatures. Therefore, the UN nuclear fuel is a viable fuel option for PT-type SCWRs.
Optimization of Production Plan and Risk Control in Qinshan III NPP
Authors: Zhou Jun
Published: April 2011

Optimized and improved measures have been implemented in Qinshan III NPP to optimize the management of routine production plan, strengthen maintenance work risk analysis, and improve the plan execution capability, which involve unified management of generation, refuelling, periodic test and maintenance plans, simplifying the defect scales and reducing interlinks of defect disposal, intensifying the assessment on plan execution and adopting performance evaluation and star rating measures.
A Comparison of Different Communication Tools for Distance Learning in Nuclear Education
Authors: Harvel Glenn , Hardman Wendy
Published: April 2011

Nuclear Engineering Education has seen a recent surge in activity in the past 10 years in Canada due in part to a Nuclear Renaissance. The Nuclear Industry workforce is also aging significantly and requires a significant turnover of staff due to the expected retirements in the next few years. The end result is that more students need to be prepared for work in all aspects of the Nuclear Industry. The traditional training model used for nuclear engineering education has been an option in an existing undergraduate program such as Chemical Engineering, Engineering Physics, or Mechanical Engineering with advanced training in graduate school. The education model was mostly lecture style with a small number of experimental laboratories due to the small number of research reactors that could be used for experimentation. While the traditional education model has worked well in the past, there are significantly more advanced technologies available today that can be used to enhance learning in the classroom. Most of the advancement in nuclear education learning has been through the use of computers and simulation related tasks. These have included use of industry codes, or simpler tools for analysis of the complex models used in the Nuclear Industry. While effective, these tools address the analytical portion of the program and do not address many of the other skills needed for nuclear engineers. In this work, a set of tools are examined that can be used to augment or replace the traditional lecture method. These tools are Mediasite, Adobe Connect, Elluminate, and Camtasia. All four tools have recording capabilities that allow the students to experience the exchange of information in different ways. The students now have more options in how they obtain and share information. Students can receive information in class, review it later at home or while in transit, or view/participate it live at a remote location. These different options allow for more flexibility in delivery of material. The purpose of this paper is to compare recent experiences with each of these tools in providing Nuclear Engineering Education and to determine the various constraints and impacts on delivery.
Advanced Dryout Correlation for BWR Fuel
Authors: Paramonov D. , Adamsson C.
Published: April 2011

Each BWR fuel design requires a method to predict its dryout performance in order to be licensed. Presently, the assessment of dry-out risk is based on empirical correlations, which sometimes results in inaccurate or non-physical predictions in certain portions of operational space. This poses a number of limitations as plant operators seek to extract additional value from the fuel through more aggressive operation strategies. A new form of BWR dryout correlation is developed. Accuracy of predictions outside of experimental data range is increased by employing a non-linear correlation form and the transformation to axial power profile, which is based on physical considerations. Proper qualitative behavior is assured by the correlation form itself rather than values of regression coefficients.
Thermalhydraulics and Safety Programs in Support of the CANDU SCWR Design
Authors: Leung L. K. H.
Published: April 2011

An extensive thermalhydraulics and safety program has been established in Canada to support the research and development effort for the CANDU supercritical water-cooled reactor design. This program focuses mainly on establishing the research capability at various Canadian universities, and includes several collaborative projects with Chinese universities to expedite the pre-conceptual design. Several test facilities in support of the supercritical heat transfer research are described. Selected results are presented from analytical studies.
Integrator Wind-Up Effect of Recirculation Flow Control System During Large Load Demand Change for Lungmen ABWR Plant
Authors: Ma Shaoshih , Shih Chunkuan , Yuann Yngruey , Wang Jongrong
Published: April 2011

The transient of Large Load Demand Change imposed on the Master Controller of the Recirculation Flow Control System is studied in this paper. The simulation performed in this study used the Core Design and Safety Analysis Package, which includes a lattice calculation code, CASMO-4, a three-dimensional core simulation code, SIMULATE-3, an interface code, SLICK, and a thermal hydraulic code, RETRAN. In the analysis the initial condition of 100% rated power/85% rated core flow was selected as base case, and the load demand was first manually ramped down from 100% to 75% (25% change) at the rate of 1%/sec and then ramped up at the same rate to the original level 150 seconds later. This paper presents the wind-up phenomenon observed in the output of the Proportional Integral (P/I) element when a large load demand change is imposed on the Master Controller of the RFCS. This wind-up phenomenon will cause an unnecessary time delay in response to the change to restore the Master Controller output to above the effective level. The sensitivity studies were conducted to get more insight into the wind-up phenomenon, which included: (1) load demand change of 20% level, (2) different initial core flow, and (3) different gain setting of P/I Controller. The analysis results show that the wind-up phenomenon can be avoided for the load demand change less than 20% or the higher initial core flow than 85% rated. Furthermore, the base setting of the P/I element of the Master Controller is appropriate to prevent the neutron flux from exceeding the scram setpoint with at least 10% margin during the ramp-up stage of load demand change.
Crack Growth Monitoring by Strain Measurement of the External Surface of Cracked Pipes: A Feasibility Study
Authors: Kamaya Masayuki , Miyoshi Koji
Published: April 2011

The cracks detected by in-service inspection are not always removed when they are not hazardous according to fitness-for-service evaluations. In such cases, it is important to monitor the growth of these cracks in order to assess the validity of their integrity assessment. However, due to the limitation of the accuracy of size determination by ultrasonic testing, it is difficult to know how much the cracks have grown since the previous measurement. In this study, a method for crack growth monitoring during plant operation is proposed. When a pipe is deformed elastically due to internal pressure, the strain at its external surface increases. By measuring the change in strain for the outside of the cracked pipe continuously, it is possible to determine how much the crack size changes. Elastic finite element analyses were performed for cracked pipes under internal pressure. From the analyses under various crack sizes, the accuracy and resolution of the proposed monitoring method were evaluated. It was revealed that the method could detect crack growth of less than several hundred micrometers.
Analysis for Inverse Flow in Moderator Channels of Supercritical Water Cooled Reactor
Authors: Fu Shengwei , Xu Zhihong , Yang Yanhua
Published: April 2011

This paper mainly talks about supercritical water cooled reactor (SCWR) fuel assembly which has special character such as moderator channels, two flue pin rows between the moderator channels and the thermal zone of fuel bundle. SABER (Sub-channel Analysis Based on Extended pressure Range) is used for sub-channel analysis of the fuel assembly. It is a self-developed sub-channel thermal-hydraulic analysis code. The code was originally developed to analyze SCWRs. After adding an extended computational cell structure and some new boundary conditions, SABER can simulate the complex flow (the opposite flow direction between moderator channels and coolant channels) in the fuel assembly. SABER can also compute heat exchange between moderator channels and cooling channels. The phenomenon of inverse flow occurs in the moderator channels, was found by sub-channel analysis. The phenomenon is caused by the different heating situation around moderator channels. Some sensitivity analysis of this phenomenon was analyzed in the paper.
Assessment of ODS Ferritic Alloy for Advanced Nuclear Energy System
Authors: Zhou Zhangjian , Li Ming , Liao Lu , He Pei , Xu Yingli
Published: April 2011

Development of cladding materials which can work at high temperature is crucial to realize highly efficient and high-burnup operation of Generation IV nuclear energy systems. Oxide dispersion strengthened (ODS) steels are the most promising class of materials with a potential to be used at elevated temperature under severe corrosion and strong neutron exposure environment. ODS ferritic steels with Cr content of 12–18% were designed and fabricated through the mechanical alloying (MA) route. The characterization of ODS ferritic steels were conducted for evaluation their potential applications for advanced nuclear energy system. Mechanical properties were measured at room temperature and high temperature. High-temperature (700 °C–1000 °C) oxidation resistance were carried out using a muffle. Microstructures of the oxidation layer were observed and compared.
Computational Analysis of Core Catcher Behavior During Corium Relocations From Reactor Vessel
Authors: Astafyeva Vera , Bezlepkin Vladimir , Sidorov Valery , Semashko Sergey , Kukhtevich Vladimir
Published: April 2011

The equipment of LNPP-2006 for severe accident’s managements includes the device of core melt localization–core catcher. By analogy with core catcher of the Tainvan NPP in China, the LNPP-2006 core catcher has been developed based on the crucible concept that combines corium retention inside the vessel (passive water cooling of metal surfaces that form the boundary of the corium localization zone) and the control of physical and chemical properties of the corium (by the use of sacrificial material). By present time design works in which the basic design parameters of the core catcher are defined have been performed. It was allowed to execute the verification calculations of core catcher processes taking into account behaviour of reactor’s cavity materials. For the analysis of core catcher severe accident’s behaviour code HEFEST-CC was used, for its adjustment three-dimensional hydraulic codes were applied. As a result of this analysis the quantity of the additional materials, which supplying from reactor’s cavity during concrete destructions under the influence of radiation was adjusted. The cooling ability of a passive water system of a melt surface has been confirmed. Introduced paper is devoted the analysis the formation dynamics and parameters of the molten core bath inside the core catcher vessel, to destruction the fracture dynamics of a thermal protection and response time the water passive feeding system on a melt surface, to definition the crisis store before heat exchange on the water-cooled vessel.
The Flowing and Heat Transfer Models of Turbulent Flow in Rolling Motion
Authors: Yan Binghuo , Gu Hanyang , Yang Yanhua , Huang Yanping
Published: April 2011

The flowing and heat transfer characteristics of turbulent flow in tubes in rolling motion are investigated theoretically. The flowing and heat transfer models of turbulent flow in rolling motion are established. The correlations of frictional resistance coefficient and Nusselt number are derived. The results are also validated with experiments. The effects of several parameters on Nusselt number are investigated. The oscillating amplitude of Nusselt number is in direct ratio with Prandtl number and rolling frequency approximately. The more the flowing velocity is, the less the effect of rolling motion on the flow is. The variation of initial phase difference between Nusselt number and rolling motion with rolling frequency is very limited.
Time Dependent Leak Rate Change in SCC Degraded SG Tubes of PWRs
Authors: Hwang Seong S. , Park Jangyul , Jung Man K. , Kim Hong P.
Published: April 2011

Primary water stress corrosion cracking (SCC) and an outside diameter SCC have occurred in the steam generator (SG) tubes of nuclear power plants in the republic of Korea and around the world. Although high corrosion resistant alloy 690 has been replacing the alloy 600 tubings, it is important to establish the repair criteria for the remaining degraded alloy 600 tubings to reassure regarding the reactors integrity, and still maintain the plugging ratio within the limits needed for its efficient operations. For assessment and management of the degradation, it is crucial to understand the initial leak behaviors and time dependent leak rate change from SCC flaws under a constant pressure. Stress corrosion cracked tube specimens were prepared at room temperature by the contact of sodium tetrathionate solution. The initial leak rate and time dependent leak rate were measured at different pressure ranges with time. Water pressure inside the tube was slowly increased in a step like manner with a designated holding time. The leak rate was calculated by dividing the amount of water by the time. A large open and long axial crack showed an increasing leak rate with time at a constant pressure, whereas small opened cracks did not show an increase in a time dependent leak rate. Under some pressures, the leak rate did not increase with the increase of pressure due to a tightness of circumferential cracks. Throughwall axial crack of 5 mm long may exhibit the leakage of action level 1 of the EPRI leakage guideline.
Predicting Real Accidents and Rare Events: The Human Contribution to Safety
Authors: Duffey Romney B.
Published: April 2011

We examine the prediction of real accident and event probability in the absence of prior data and/or with partial knowledge, when the human contribution is properly included. We now know that the major cause of all real accidents (not postulated ones) is actually the unforeseen human contribution, as an integral and inseparable part of the technological system. The real events we actually will experience or observe in our lives may be spectacular plane, train, space shuttle or stock market crashes. In every case, they are unexpected occurrences, they seemingly appear randomly, and how often they happen, or the rate of such events, covers the whole spectrum from frequent to rare. Because so-called rare events do not happen often, they are also widely misunderstood and do not follow the expectations or the same “rules” governing many or frequent events, and are always due to some apparently unforeseen combination of circumstance, conditions, and combination. Usually in safety analysis, a distinction is made between “probabilistic” safety analysis (PSA), based on examining so-called risk dominant accident sequences, and “deterministic” safety analysis (DSA). Intended to be complementary, PSA provides insights into risk scenarios and allowing numerical estimation of outcomes for transients, such as loss of offsite power (LOOP) or station blackout (SBO), and the resulting core damage frequency (CDF) or large early release frequency (LERF) with some estimated uncertainty in the calculated probabilities of occurrence. In contrast, the DSA provides a standard set of stylized events, such as large breaks (LOCA) and transients (ATWS), as a means of setting safety margins and design criteria, as also proposed in Theofanous’s ROAMM, where extremes of knowledge are postulated as a test of the robustness of the design and safety systems. These methods can produce statements of margins and uncertainties, and converge in the area known as “risk informed regulation” (RIR), where the insights gained are proposed to derive limiting Farmer-type “tolerable risk” boundaries or frequency-consequence (F-C) curves. Conversely, real accidents are often unknown sequences, with no priors or precursors, and/or include possibly unforeseen initiators (for example, undetected pressure vessel corrosion) and the key role of the human. In this paper, we address the question of the quantitative prediction of such real and rare events, their occurrence probability and hence the risk.
The Effect of Load Reduction on Crack Initiation Behavior of Hydrides From Flaws in Zr-2.5Nb Pressure Tube Material
Authors: Cui Jun , Shek Gordon K.
Published: April 2011

Flaws in Zr-2.5Nb alloy pressure tubes in CANDU® nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC). DHC is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth of the hydrided region, and fracture of the hydrided region at the flaw tip. One scenario of crack initiation is that the flaw-tip hydrides are formed and cracked at the same stress under constant reactor operating pressure. This is known as crack initiation under constant-load condition. Another scenario of crack initiation is that the flaw-tip hydrides are formed at the operating pressure and then cracked during a transient over-pressure. This is known as crack initiation under overload condition as the hydrides are subjected to a stress higher than the hydride formation stress. In some CANDU reactors, a 20% reduction in pressure is implemented during reactor cool-down. This paper examines the effect of pressure reduction, and hence load reduction, on flaw-tip hydride morphology and crack initiation behavior under constant-load and overload conditions. Experiments were performed on specimens of an unirradiated Zr-2.5Nb pressure tube, with 57 wt. ppm hydrogen concentration. The specimens contained machined V-notches with a root radius of 0.015 mm to simulate service-induced debris fretting flaws. The results indicate that the 20% load reduction increases the threshold stresses for crack initiation under constant-load and overload conditions. Finite element stress analyses were performed to determine the notch-tip stress distribution under constant-load and 20% load-reduction conditions. The load reduction lowers the notch-tip peak stress and shifts its location away from the notch surface. This is consistent with the notch-tip hydride morphologies observed using optical and scanning electron microscopy.
Preliminary Analysis of Seismic Shutdown PSA of a PWR plant
Authors: Kondo Keisuke , Fujimoto Haruo , Yamashita Masahiro , Nishio Masahide
Published: April 2011

Shutdown seismic PSA was performed for the plant cold shutdown condition. Normally, automatic ECCS signal is blocked after the Permission Signal is indicated during the plant cooldown process of PWR. In response to this, PSA model including manual actuation of ECCS signal on the postulated LOCA condition during cold shutdown state was developed based on the actual plant Emergency Operation Manual, which consists of several steps, such as, 1) LOCA event diagnosis, 2) isolating RHR system, 3) manual actuation of ECCS signal, 4) switching RHR to Low Pressure Injection mode. This PSA model with evaluated human error rates, was quantified demonstrating that risk level for the cold shutdown condition is in the acceptable range, however, suggesting some room for model improvement considering more realistic manual ECCS actuation procedure depending on the size of the LOCA.
Comparison of Publicity Materials Pertaining to the Selection of High Level Radioactive Waste Repository Sites
Authors: Kanzaki Noriko , Okamoto Koji , Wada Ryutaro
Published: April 2011

The high level radioactive waste was decided to the geological disposal in Japan. In addition, operating body and capital management body were decided. However, the place in the repository site has not been decided yet. The selection process in the waste repository is already advanced in Finland and Sweden. In England, some municipalities have expressed interest in HLW. The word used in publicity materials is analyzed, systems in each country are compared, and some good idea for Japanese HLW repository site selection process is studied.
Study on Pipe Wastage Mechanism by Liquid Droplet Impingement Erosion II (Measurement of Erosion Rate)
Authors: Weng William Khoo Chong , Higashi Yuma , Narabayashi Tadashi , Sugiyama Kenichiro , Ohshima Hiroyuki , Kurihara Akikazu
Published: April 2011

Evaluation of wastage rate for nuclear power plant piping is important in maintaining plant reliability and safety. The two commonly well known causes of pipe wastage are liquid droplet impingement erosion (LDIE) and flow accelerated corrosion (FAC). This study is to evaluate the wastage rate of pipe wall and to measure the loss of mass caused by LDIE. A high speed rotating disk and water jet is used to perform LDIE phenomena. In this study, the liquid droplet impinging tests were conducted with the test piece mounted on the high speed rotating disk which crosses thin water jet. The amount of the wastage by LDIE was evaluated by changing the rotational speed, test time and test piece materials. After the experiment, the result of erosion was investigated by observing the surface of the test piece using a digital microscope, and photo was taken. A method of converting the photo into data is used. In this study, LDIE is considered as the main cause of test piece mass difference before and after the experiment. Therefore, a method is designed to evaluate the mass difference. In this method, the difference in volume of the material is calculated. The loss of mass is obtained by multiplying the difference in volume and density of the material. Then, 2-dimensional and 3-dimensional graphs of the test piece are drawn. The analytical code TRAC is used to analysis the liquid droplet velocity in Advance Boiling Water Reactor (ABWR) drain pipe.
A Decision Support System for the Off-Site Emergency Protective Actions Based on the OILs for Nuclear Emergencies
Authors: Ling YongSheng
Published: April 2011

Deterministic health effects can be prevented and the risk of stochastic health effects can be reduced by taking protective actions before or shortly after a release. These actions must be based on plant conditions and then refined subsequently based on environmental measurements. Operational intervention levels (OILs) are some calculated values (e.g., ambient dose rate or radionuclide concentration) measured by instruments or determined by laboratory analysis that correspond to a GIL or GAL. Through the use of the OILs, the environmental data are assessed primarily, which are quantities directly measured by the field instrument. Default OILs have been calculated in advance on the basis of the characteristics of severe reactor accidents. These default OILs are used to assess environmental data and take protective actions until sufficient environmental samples are taken and analyzed to provide a basis for their revision. This approach allows data to be quickly evaluated, and decisions on protective actions to be promptly made. A decision support system of the off-site emergency protective actions based on the OILs for nuclear emergencies was discussed in this paper. The system accesses the environmental data through the default OILs and the revisions of OILs. It is applied to Daya Bay Nuclear Power Plant. In the early release, according to the characteristic of the plant, the system provides the approach to calculate the default OILs based on the accident source terms described in Reactor Safety Study of USA. Also some real factors are considered, including the meteorological parameters. When sufficient environmental samples are taken and analyzed to provide a basis for their revision, the default OILs can be revised or recalculated by them. In the entire emergency planning zone, the environmental data will be assessed through the use of those OILs to provide the advice of protective measures.
Dr. Mainte: An Integrated Simulator for Optimization of Maintenance Strategies of Light Water Reactors
Authors: Isobe Yoshihiro , Sagisaka Mitsuyuki
Published: April 2011

Dr. Mainte, an integrated simulator for maintenance optimization of LWRs (Light Water Reactors) has been developed based on PFM (Probabilistic Fracture Mechanics) analyses. The concept of the simulator is to provide a decision-making system to optimize maintenance activities for representative components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability and economic efficiency.
Prediction of the Critical Heat Flux in Subcooled Flow Boiling in Round Tube Using an Improved Mechanistic Model
Authors: Yu Na , Zhang Yu
Published: April 2011

Among the existing CHF models, the bubble crowding model and the liquid sublayer dryout model have been well accepted for subcooled flow boiling. But both of the two models couldn’t give explanation about some details in the boiling crisis phenomenon according to photographic result. The aim of the present paper is to provide an improved synthesized model containing the characteristic of the above two models and then to give a comprehensive explanation about CHF. In the present model, the conservation equations of mass and energy are solved to derive the CHF formula. The length and velocity of the vapor blanket and the thickness of the liquid sublayer are needed. The quality and void fraction in bubble region and the core region are calculated by a homogeneous assumption. The vapor blanket length is thought to be equal to the Helmholtz wavelength and it is obtained from several parameters in the bubble region. The velocity of the vapor blanket is connected to the flow velocity of the bubble layer. The thickness of the sublayer is determined by a force balance on the vapor blanket, which is also related to the condition of the bubble region. About 1100 experimental points have been selected to verify the proposed model. Comparison between the predictions by the proposed model and the experimental result shows a good agreement that more than 90% of these data are predicted within ±20%.
Materials Research in Support of SCWR Development: Current Areas, Gaps and Needs
Authors: Zheng Wenyue , Guzonas Dave , Brady D. , Cook W. , Luo J.
Published: April 2011

International efforts on materials selection and development related to supercritical water-cooled reactors (SCWRs) have produced a considerable amount of data in the open literature [1], The majority of these data are on aspects of materials properties such as corrosion, stress corrosion cracking, creep, irradiation damage as well as microstructural degradation under various exposure conditions. These prior efforts are helping guide the current selection of candidate alloys for further, longer-term evaluation. As continuing research on the SCWR advances, gaps and limitations in the published data are being identified. In terms of corrosion properties, these gaps can be seen in several areas, including: 1) the test environment, 2) the physical and chemical severity of the tests conducted as compared with likely reactor service/operating condition, and 3) test methods used. While some of these gaps can be filled readily by the current research projects, in particular those occurring in Generation IV International Forum (GIF) member countries, others require further advances in our understanding of material-environment interactions involving supercritical water. Gaps in advanced test facilities for future research are also becoming evident. Future needs for materials development and suggestions for expanded international collaborations to link with groups working on materials for advanced fossil-fired supercritical water power plants, as well as other GEN IV and fusion reactor designs are summarized.
Research of Nuclear Emergency Management Information System in Guangdong Province
Authors: Ling YongSheng
Published: April 2011

Nuclear off-site emergency response for nuclear power plants in Guangdong province is a complicated process, which is finished by many emergency response organizations. In a nuclear emergency, the source, format, content, take-over means and processing means of the information passed among these organizations are different. The aim of nuclear emergency response is to protect public and environment by taking some proper urgent protective action. The essential information from the original information is necessary for decision-making in the processing of nuclear response emergency, thus the key of emergency response is incepting and analyzing the information. An effective method of processing information is using a computer-based information system to aid the traditional handwork means. Nuclear Emergency management Information System in Guangdong Province is such systems, which covers preparedness, response and recover phase. At first, this paper analyzed the system functions according to the responsibilities of the emergency organizations in Guangdong province. Then the system frame is decided, which is composed of operation, information management, file processing, consequence assessment, operational intervention level (OIL) computing subsystem, geographic information system and a system database. A useful method for decision is applying the OILs in consequence assessment, and default OILs need be revised in a real nuclear emergency. Developing OIL computing subsystem is helpful to make decision with measurement result directly.
A Convenient Multi-Model Simulation Technique for Nuclear Reactor System
Authors: Peng Jinfeng , Zhao Fuyu , Liu Cheng
Published: April 2011

In the field of system and control, the treating for nonlinear system is always the hot issue in present research. The multi-model method is an important method for solving the nonlinear problem. In traditional solution, the performance of multi-model simulation is not convenient enough. In this paper, a simple, visual and convenient simulation technique is introduced. The toolbox “Simulink” which is derived from the software “Matlab” is used for simulating dynamic process. As a new original attempt, the transfer functions of system models are transformed to module charts. According to the module charts, the multi-model system module is built. The advantage of this technique is that: the coefficients in transfer functions are online controllable. Then the switch and connecting rules between sub-models are introduced. The switch and connecting rules are designed according to the actual case. All of above elements will be designed to module in Simulink workspace. Finally, an illustration is taken. With using the method recommended in this paper, the operation of the simulation is visual, simple and convenient. The technique recommended in this paper can also be suitable for other fields which include nonlinear system.