bluuebunny/crossref_metadata_2025 huggingface dataset:
Employee share ownership and investment concentration: which employee shareholders fail to diversify?
Authors: Pendleton Andrew
Published: February 2010

Using data from a survey of employees in three UK companies, the article examines concentration of savings in employer shares. It is found that a substantial minority of employees hold most of their savings in employer shares despite the risks. The article examines the distribution and determinants of portfolio concentration. It is found that portfolio concentration is more likely to be found among men, those on higher incomes, and those who are in early middle‐age. Regression analysis indicates that organisational commitment and the duration of participation in the share ownership plan also affects the probability of having a concentrated portfolio. There is no clear evidence that those with substantial concentrations of employer shares take active steps to monitor their investments more closely than those with smaller shareholdings.
B‐Type Natriuretic Peptide: Application in the Community
Authors: McDonald Ken , Dahlström Ulf , Aspromonte Nadia , Jourdain Patrick , Maisel Alan , Mueller Christian , Valle Roberto
Published: February 2010

Natriuretic peptide assessment has represented a significant advance in the management of heart failure. In a syndrome in which clinical symptoms and signs can be either nonspecific or absent, the presence of a reliable biomarker to aid diagnosis, assess prognosis, and potentially guide treatment and aid in prevention of this syndrome has represented a significant advance. The following review will outline established and potential new roles for natriuretic peptide assessment in the community. Congest Heart Fail. 2008;14(4 suppl 1):12–16. ©2008 Le Jacq
Emergency Department Perspectives on B‐Type Natriuretic Peptide Utility
Authors: Peacock W. Frank , Mueller Christian , DiSomma Salvatore , Maisel Alan
Published: February 2010

In the emergency department (ED), patients do not present with a diagnosis. Rather they manifest a symptom for which the physician must determine the most probable cause. It is the sorting of possibly acutely fatal diagnoses that is unique to the practice of emergency medicine. Unfortunately, the historically available rapid diagnostic tools lack both sensitivity and specificity. This creates risk for the acutely ill patient in whom either a delayed or erroneous diagnosis can be associated with a startling increase in mortality. B‐type natriuretic peptide (BNP) assessment in the ED is a valuable adjunct for increasing the accuracy of the initial clinical impression. This allows early outcome‐improving interventions to proceed with greater certainty. Finally, as a prognostic marker indicative of illness severity, knowledge of BNP levels can improve the accuracy of disposition decisions. Congest Heart Fail. 2008;14(4 suppl 1):17–20. ©2008 Le Jacq
Up‐and‐Coming Markers: Myeloperoxidase, a Novel Biomarker Test for Heart Failure and Acute Coronary Syndrome Application?
Authors: Sinning Christoph , Schnabel Renate , Peacock W. Frank , Blankenberg Stefan
Published: February 2010

Myeloperoxidase (MPO) is a mammalian enzyme responsible for generation of hypochlorite. The advantage of myeloperoxidase for use as a biomarker in the setting of heart failure and acute coronary syndrome is the early increase of MPO concentration in response to the acute event. In the setting of heart failure the reported independency of coronary artery disease and general inflammation, as indicated by MPO concentration in comparison to other inflammatory markers or in subgroups of patients with ischemic and non‐ischemic cardiomyopathy, has to be highlighted. In terms of ACS, inclusion of MPO into a multiple marker strategy might add to enhance diagnosis and therapy decision making. Therefore, MPO is a biomarker worthwhile of further evaluation in the setting of cardiovascular disease. Congest Heart Fail. 2008;14(4 suppl 1):46–48. ©2008 Le Jacq
Experience With Implantable Loop Recorders for Recurrent Unexplained Syncope
Authors: Brignole Michele
Published: February 2010

Knowledge of what occurs during spontaneous syncope is the gold standard for evaluation. Initially, implantable loop recorders (ILRs) were used in patients with unexplained syncope at the end of unsuccessful full, conventional work‐up. In pooled data regarding 247 patients, a correlation between syncope and electrocardiographic findings was found in 84 patients (34%); of these, 52% had a bradycardia or asystole at the time of the recorded event, 11% had tachycardia, and 37% had no arrhythmia. Presyncope‐electrocardiography correlation was observed in another third of the patients; presyncope was much less likely to be associated with an arrhythmia than was syncope. The diagnostic yield was similar in patients with and without structural heart diseases and was higher in older than in younger patients. Recent studies showed that ILR implantation can be safely performed in an early phase of the diagnostic evaluation—provided that patients at risk for life‐threatening events are carefully excluded—in the patients who have a severe presentation of syncope (because of high risk of trauma or high frequency of episodes) which can be a benefit of a mechanism‐specific therapy. Congest Heart Fail. 2008;14:7–12. ©2008 Le Jacq
The Use of Device‐Based Diagnostics to Manage Patients With Heart Failure
Authors: Germany Robin
Published: February 2010

Heart failure patients continue to increase in number and complexity, with a growing need for better management strategies. With a significant number of these patients also having implantable devices, the opportunity to treat patients utilizing the information stored in the devices is tremendous. The data available include arrhythmias, heart rates, activity levels, and new heart failure monitoring markers. Additional metrics are in development and may provide improved monitoring and treatment capabilities. Information can be accessed in a number of ways, with increasing availability either in the office or remotely. Learning how to access and utilize these data will benefit patients currently and allow the incorporation of the significant amounts of data that are currently in development. Congest Heart Fail. 2008;14:19–24. ©2008 Le Jacq
‘Getting the kids back to school’: education and the emotional geographies of the 2007 Hull floods
Authors: Convery I. , Balogh R. , Carroll B.
Published: February 2010

Major flood events almost inevitably affect education and are likely to have a lasting impact on a school and community. In this paper, based on interviews with head teachers carried out in Hull during December 2008, we focus on the 2007 flood event in Hull and the way this affected schools and pupils in the city. The paper indicates the importance of reintegrating children (and families) into community structures such as schools as soon as possible after the flood and of creating safe spaces (‘circle time’) within the school classroom for pupils to explore their flood experiences. While we have commented on the potential therapeutic value of circle time in a post‐disaster situation, we acknowledge that more research is required.
Urban pluvial flooding: a qualitative case study of cause, effect and nonstructural mitigation
Authors: Douglas I. , Garvin S. , Lawson N. , Richards J. , Tippett J. , White I.
Published: February 2010

Historically, flood risk management in the United Kingdom has mainly concentrated on river and coastal flooding, yet flooding from surface water runoff is a risk to urban areas. A comprehensive study of the causes, the impact and the consequences as well as the management of serious pluvial flooding in Heywood, Greater Manchester, in 2004 and 2006 revealed that the victims of the floods were unprepared, ill‐informed and confused as to responsibilities before, during and after the event. Householders had to rely on their insurers for loss mitigation, but the response of the insurance industry was varied and inconsistent, and there were difficulties in building in resilience after the event. In 2006, only one property was on the Office of the Water Regulator DG 5 Register on the basis of previous flooding. Thus the area falls between the responsibilities of the Local Authority (LA), the Environment Agency and the water utility. The people affected do not know whom to turn to for assistance. A way forward may be through the establishment of an overriding agency to provide a coherent voice and strategic guidance, supported by dedicated flooding experts within LA planning departments, the adaptation of buildings for flood resilience and through changes within the insurance industry.
Development of a New Spacer Grid Form to Enhance the Integrity of Fuel Rod Support and the Crush Strength of a Spacer Grid Assembly
Authors: Song Kee-nam , Lee Soo-bum , Shin Moon-Kyun , Lee Jae-Jun , Park Gyung-Jin
Published: February 2010

A spacer grid is one of the most important structural components in a LWR fuel assembly. The spacer grid, which supports nuclear fuel rods laterally and vertically with a friction grip, is an interconnected array of slotted grid straps welded at the intersections to form an egg-crate structure. Dimples and springs are stamped into each grid strap to support the fuel rods. The form of grid straps and spring form is known to be closely related with the crush strength of spacer grid assembly and the integrity of fuel rod support, respectively. Zircaloy is prevailing as the material of the spacer grid because of its low neutron absorption characteristic and its successful extensive in-reactor use. The primary considerations are to provide a Zircaloy spacer grid with crush strength sufficient to resist design basis loads especially due to seismic accidents, without significantly increasing pressure drop across the reactor core. Generally, the thickness and height of the Zircaloy grid strap have been the main design variables in order to meet the above considerations. Recently, it was reported that a dimple location is also a design variable that affects the crush strength of a spacer grid assembly. In this study, a new spacer grid form was developed in order to enhance the integrity of the fuel rod support and the crush strength of the spacer grid assembly by using a systematic optimization technique. Finite element analysis and crush strength tests on the developed new spacer grid form were carried out to check the performance enhancement compared to commercial spacer grids. The enhancement of fuel rod support was confirmed by comparisons of contact area, peak stresses, plastic deformation and etc. According to the results, it is estimated that the actual critical load enhancement of the spacer grid assembly is approximately up to 30% and the actual contact area, when a fuel rod inserted into a spacer grid cell, is more than double for the developed new spacer grid form. And also, some design variables that effect the crush strength of a PWR spacer grid assembly were classified and their effects on the crush strength were investigated by a finite element analysis and a crush strength test.
Research of Nuclear Power Plants Diagnosis Method Based on Data Fusion
Authors: Liu Yong-kuo , Xia Hong , Xie Chun-li
Published: February 2010

Data fusion is a method which suits for complex system fault diagnosis such as nuclear power plants, and is multi-source information processing technology. In this paper, the data fusion information hierarchical thinking used and the nuclear power plants fault diagnosis divided into three levels. In data level data mining method adopted to handle data and reduction attributes. In feature level three parallel neural networks used to deal with attributes reduction of data level and the outputs of three networks are as the basic probability assignment of Dempster-Shafer (D-S) evidence theory. The improved D-S evidence theory synthesizes the outputs of neural networks in decision level, which conquers the traditional D-S evidence theory limitation that cannot dispose conflict information. The diagnosis method is tested through using correlation data of document. The test results indicate that the data fusion diagnosis system can diagnose nuclear power plants faults accurately and the method which has a certain applicable value in use.
Seismic Isolation of Nuclear Structures by a Non-Oscillating System SERB Type
Authors: Serban Viorel , Panait Adrian , Androne Marian , Ciocan George Alexandru
Published: February 2010

The paper presents the advantage of non-oscillating systems — SERB type — for the seismic isolation of buildings compared with other solutions for seismic protection of the buildings which are now employed worldwide. The non-oscillating isolation devices — SERB type — have a capsulated structure and are capable to overtake large compression and tension loads on vertical direction with very small deflection and cut off the seismic action in horizontal direction. The stiffness of the devices is zero for the translations on any direction in horizontal plane upto a pre-set displacement which usually, is between 0.6–0.8 of the maximum seismic soil displacement. After having reached that value, a non-linear increase of the device stiffness is occurring on the horizontal direction in order to provide a non-shock limitation of some random seismic displacements higher than the maximum displacement accommodated by the device. The motion inside the device is controlled all time by the dynamic friction and restricted by nonlinear stiffness and damping material after a pre-set displacement. On vertical direction, the stiffness of the devices is big in order to avoid the additional stresses on the structural elements of the isolated suprastructure, due to the differential settlement of the devices on the suprastructure permanent loads that could appear. The system provides very small deflection of the device on vertical direction with high damping. On request, SERB isolators can be so fabricated to have the desired stiffness on vertical direction. Standard SERB devices 960×960×230 are capable to overtake a permanent compression load of 3000KN and tensile load of 1500KN, respectively, over which dynamic loads of similar values may overlap. The maximum seismic acceleration transferred to the isolated structure on horizontal direction is ranging between 0.03g and 0.05g depending on the friction coefficient. The maximum relative displacement between the isolated suprastructure and the infrastructure embedded into the ground, is about 3 times smaller than the maximum seismic displacement showed by the oscillating systems employed today and smaller then the maximum seismic soil displacement. In series SERB device 960×960×230 allows for a maximum seismic displacement of 230mm. An application of the non-oscillating SERB devices 960×960×230 as the solution to seismically isolate the heavy water detritiation system building at Cernavoda NPP, is presented.
Approach on Component Selection for Ageing-Trend Analysis Within PSA Models
Authors: Poghosyan Shahen , Amirjanyan Armen , Malkhasyan Albert
Published: February 2010

The major advantage of PSA is the possibility of in-depth qualitative and quantitative analysis of NPP actual configuration with definition of factors introducing a significant contribution to the general risk of reactor core damage. However main lack of the PSA current models is neglect of equipment ageing effects. Neglecting of ageing effects in PSA could lead to incorrectness of risk profile and influent on risk-informed decision making process. To solve this issue incorporation of ageing aspects into PSA models for Armenian NPP Unit 2 was initiated. Implementation of ageing trend analysis for all PSA components is insuperable effort, so the first step of the analysis is component selection activity. This paper is addressing the approach on component selection for ageing-trend analysis within PSA models. Presented approach is based on ageing effect and risk importance data. The procedure was developed and implemented in the framework of ageing aspects incorporation into PSA level 1 model for Armenian NPP Unit 2.
Modelling of Material Properties Influence in Fuel-Coolant Interaction Codes
Authors: Ursˇicˇ Mitja , Leskovar Matjazˇ , Kljenak Ivo , Mavko Borut
Published: February 2010

The melt droplets crust formation modelling, which is used in current FCI (fuel coolant interaction) codes, is rather basic. In the paper the development of the melt droplet heat transfer model, which enables the treatment of the material properties influence on the steam explosion is presented. The model is enough complex to correctly predict the crust development during the melt droplets cooling in the premixing phase. At the same time the model is enough simple that it can be practically implemented into FCI codes and is thus being an optimal model for FCI applications. Fragmentation criteria are derived in order to take into account the influence of the formed crust on the steam explosion process. The derived criteria are based on experimental results and the thin plate approximation. To enable the use of the model and the fragmentation criteria in FCI codes with Eulerian formulation, adequate transport equations for model parameters are given.
On the Use of Binary Importance Decision for Risk-Informed Categorization of SSCs
Authors: Zeng Wei , Yu Hongxing
Published: February 2010

The categorization of Structures, Systems and Components (SSCs) is one of the foundations in the design of the nuclear reactor. The regulation requirements based on the current deterministic categorization are over-conservative in some aspects; on the other hand, some requirements for the nonsafety-related components are too loose. In order to make the requirements reasonable, the risk-informed significance categorization of SSCs has been presented. Risk-informed significance categorization (RISC) is to categorize structures, systems, or components (SSCs) of a nuclear power plant (NPP) into two or more groups, according to their significance using both probabilistic and deterministic insights. The SSCs are quantitatively categorized by their importance measures; Fussell–Vesely (FV) and Risk Achievement Worth (RAW) are widely used. But in conventional methods for the RISC, first, a component will be categorized as significance once the value of FV or RAW is over the thresholds. So the significant components are treated equally regardless of the difference between FV and RAW, that is not suitable. Second, the component RAW derived from the sum or maximum of the basic events is not realistic. Third, the categorization threshold for FV is not uniform, different reactors have the different thresholds. The three key problems will be researched in this paper, the quartered way will be presented base on the discussion about combination of the two importance measures (binary importance decision). And then through transferring the additivity from FV to RAW, the realistic component RAW derived from FV will be got. Finally, according to the relationship between FV and RAW, the threshold for FV will be gained similar to RAW. In the study, the author will use the new method to make a practice on the Daya Bay Nuclear Power Plant, 43% components of the low pressure safety injection system which are safety-related are categorized as low significant; 2.6% components which are nonsafety-related of the auxiliary feed water system are categorized as significant.
Water-Ingress Accident of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor
Authors: Yanhua Zheng , Lei Shi
Published: February 2010

Water-ingress accident, caused by the steam generator heating tube rupture of a high temperature gas-cooled reactor, will introduce a positive reactivity to lead the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Increase of the primary circuit pressure may result in the opening of the safety valve, which will cause the release of radioactive isotopes and flammable water gas. The analysis of such an important and particular accident is significant for verifying the inherent safety characteristics of the pebble-bed modular high temperature gas-cooled reactor. Based on the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the design basis accident of double-ended guillotine break of a heating tube has been analyzed by using TINTE, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature and primary loop pressure, the graphite corrosion inventory, the water gas releasing amount, as well as the natural convection influence under the condition of the failure of the blower flaps shut down, have been studied in detail. The calculation result of the design basis accident indicates that, the maximal possible water ingress amount is less than 600 kg and the maximal fuel temperature keeps far below the design limitation of 1620°C. The result also shows that the slight amount of graphite corrosion will not damage the reactor structure and the fuel element, and there is no potential explosive risk caused by the opening of the safety valve.
A Lifecycle Management Program for NPP Turbine Balance of Plant
Authors: Kawai Toshinari , Yamakami Katsuhiko , Hiraoka Satoshi , Manabe Jun
Published: February 2010

A lifecycle management program, for turbine balance of plant of light water reactor units, was proposed and implemented from the view point of system and equipment supplier to secure the high availability factor throughout the long intended residual lifetime. The program consists of unit surveillance analyzing operation and inspection data, degradation assessment for the equipment and prospecting for the future by appropriate measures to address the issues over the units based on both technical and economical criteria. The surveillance revealed the issue that the generating power would be adversely affected by main steam pressure reduction due to the scale adhesion to SG tubes. The prospect for the unit future was presented as an alteration of the water treatment to HAVT accompanied with optimum design for the replacement of the auxiliary heat exchangers of MSR and feedwater heaters.
The Limit Exceedance Factor Importance Measure: An Application to the Prism Reactor Design
Authors: Johnson B. C. , Apostolakis G. E.
Published: February 2010

Probabilistic Risk Assessment has several different standard importance measures for structures, systems, and components (SSCs). These importance measures are Risk Achievement Worth (RAW), Risk Reduction Worth (RRW), and Fussell-Vesely (FV). In particular, RAW and FV have been used to identify SSCs of low risk significance. However, when designing a conceptual reactor with a safety goal in mind such as those proposed in various technology neutral frameworks, these importance measures are quite crude. RAW is defined such that a component is always in the failed state, and a designer may not be looking to remove a system but to simplify it to improve economics. The Limit Exceedance Factor has been developed as a more informative importance measure when there is a goal in mind. It is defined as the factor by which the failure probability of a component may be multiplied by such that the end state (e.g. core damage) frequency exceeds a limit. With a living PRA available throughout the design process, it could allow a designer to know how much room there is for possible simplification in redundant systems. Alternatively, in the case where a system does not meet the frequency limit it can show which systems might be ideal targets for improvement to reach the limit. For a sodium-cooled fast reactor design, using the goals and limits outlined in NUREG-1860 (Technology Neutral Framework), one end state that must stay below a threshold is a very large release due to an energetic scenario. Several SSCs can be identified as risk insignificant with RAW values below two. The other SSCs tend to have very high RAW values on the order of one thousand. When the LEF is applied to the SSCs that seem to be very important using traditional importance measures, we see that some may actually have a fairly wide margin from their cited failure probabilities and still maintain the frequency goal.
Characteristics and Operation of Neutron Transmutation Doping in HANARO Reactor
Authors: Park Sang-Jun , Kim Myong-Seop , Kang Ki-Doo , Lim In-Cheol
Published: February 2010

The Neutron Transmutation Doping (NTD) of silicon is a method to produce a high quality n-type semiconductor. The NTD technology makes it possible to dope a silicon ingot with an extremely uniform dopant distribution. Korea Atomic Energy Research Institute (KAERI) has been providing the NTD service for 5 and 6 inch silicon ingots from 2003 and 2005 respectively at using HANARO. Coping with recent market demands for a silicon semiconductor by NTD, an additional irradiation facility which has a potential for the 8 inch silicon ingots was developed and the test irradiation is under way from the end of 2008. It is expected that a full scale production of 8 inch NTD-Si will be started from early 2009 and then the total capacity will be increased to 50 tons per year. This paper describes the general characteristics and operation of the NTD facility in HANARO.
Numerical Simulation of the Behavior of Fission Products in the Primary Circuit of the VVER During the LOCA Severe Accident
Authors: Tsaun S. V. , Bezlepkin V. V. , Kiselev A. E. , Potapov I. A. , Strizhov V. F. , Zaichik L. I.
Published: February 2010

The methods and models for the analysis of the radiological consequences of the design basis and severe accidents in a Nuclear Power Plant (NPP) are presented in this paper when using the system code SOCRAT. The system code SOCRAT/V3 was elaborated for a realistic analysis of radiological consequences of severe accidents in a NPP. The following models of the fission products (FP) behavior are included into the code SOCRAT/V3: (i) the condensation and the evaporation of the FP in the gaseous phase and (ii) the sedimentation, the evaporation, and the coagulation of the aerosol-shape FP. The latter processes are governed by gravity, Brownian and turbulent diffusion, thermophoresis, turbophoresis and so forth. The behavior of the FP during the loss-of-coolant accidents (LOCA) is presented to demonstrate the possibilities of the code SOCRAT/V3. The main stages of the accident (the core dryout, the core reflooding, the core degradation, the hydrogen generation, the FP release, etc.) are described. Corresponding estimations of the mass, activity, and decay heat of the suspended, settled and released into containment the FP (Xe, Te, Cs, CsI, Cs2MoO4, and so forth) are represent as well.
Reactor Vessel Closure Head Drop Analysis: Sensitivity Study on the Effects of Representing Nonlinear Behavior in the Closure Head Assembly
Authors: Castillo William C. , Remic Joseph M. , Demetri George J. , Marx Frank J. , Roarty David H.
Published: February 2010

Nuclear power plants need to safely and efficiently remove their reactor vessel closure head assembly during plant outages. This is accomplished by lifting the closure head assembly out of the reactor vessel cavity and placing it on the closure head stand. In order for nuclear power plants to remove their closure head assembly, the United States Nuclear Regulatory Commission has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis [1]. The goal of head drop analyses is to qualify the maximum drop height in air per plant procedures. A significant percentage (greater than 30%) of the closure head assembly’s mass is comprised of components attached to the top of the head (such as: lifting fixtures, a missile shield, air cooling systems, and control rod drive housings). The analytical consideration of large deflection, plastic deformation, and local failure of these components can potentially change the energy imparted to the vessel during impact due to their energy-absorbing capacities during the drop event. This paper contains a sensitivity study to determine the benefits of modeling closure head assembly components, using nonlinear structural behavior. The guidelines of Nuclear Energy Institute Initiative NEI 08-05 [2] are followed for this study.
Vulnerability of Nuclear Power Plants to Voltage Sag Induced by Disturbances in the Transmission Grid: The Concept of Zone of Vulnerability for Voltage
Authors: Abi-Samra Nick , Del Rosso Alberto , Rahn Frank
Published: February 2010

Nuclear plants are particularly sensitive to events on the grid that may lead to undervoltage on the auxiliary or safety buses. Momentary voltage dips can cause separation from offsite power and operation of emergency diesel generator. Grid disturbances that occur in certain areas or “zones” in the surrounding network, may affect the operation of the nuclear plant, while disturbances from outside these zones may be cause no threat to the nuclear plant. A zone inside which a nuclear power station would be vulnerable to events on the surrounding grid can be defined as the Zone of Vulnerability (ZoV). Different types of ZoVs can be defined depending on the nature of the vulnerability being considered. This paper deals specifically with the Zone of Vulnerability associated with the risk of voltage sags in the safety and auxiliary for safety and auxiliary buses of nuclear plants induced by faults in the power grid. The paper first introduces the concept of ZoV_v and its importance for nuclear plants security. A methodology for determining the ZoV_v is then described.
Detailed Engineering Customization: A Cost and Time Effective Approach in the Manufacturing of Steel Structures
Authors: Janjua Ajaz Bashir , Khalil M. Shahid , Akbar Sajjad
Published: February 2010

Accurate, reliable and cost effective design of steel structures in heavy & medium engineering industrial sector plays a key role in the growth of industrial development. Objective is to get awareness about the difficulties faced by engineers / managers for reducing time to minimum level for preparation of drawings and other detailing work. New concept of detailed engineering techniques is introduced, which guides, how to customize design detailing stage and how to overcome (i-e optimize) delays to meet final production targets of the product in an Heavy Duty Engineering company specialize in the manufacturing of huge and massive steel structure frames. Design engineers must understand basic tools, methods, latest techniques which will be used for minimizing cost and time effect in the production of steel structures. Extensive knowledge of AutoCAD, AutoLISP for design and customization technique is required for the production drawing as output for shop floor. For sustainable engineering development, it becomes utmost important to utilize detailed engineering customization technique. This approach mainly governs on theoretical design approach, practical / on-the-job experience and computer aided design techniques. Use of all above approaches can lead to the successful achievement of efficient engineering optimization techniques / procedures. Because of complex and complicated nature of work of steel structures in Heavy Industries, it is essential to customize such work. Therefore it is suggested to use such methods for the mass production of steel structures on time and hence lot of time could be saved in heavy duty steel industries.
Verification of Modified SCDAP/RELAP5 Against CANDU Fuel Deformation Experimental Results
Authors: Mladin Mirea , Dupleac Daniel , Prisecaru Ilie
Published: February 2010

Two new models for treating the fuel channel during the early phase of a severe accident in CANDU reactors were recently created and tested in SCDAP/RELAPSIM/MOD3.4. These treat the deformations phenomena and the possible relocation of metallic melt mixture inside the fuel channel, respectively. The horizontal relocation model consists in a series of modifications in the original Relocation-Flow-Solidification (LIQSOL) model in SCDAP code. The purpose of the paper is to apply the code to an experiment dedicated to CANDU fuel bundle behavior during degraded cooling and to attempt a validation of the model. The work generates and discusses the results of the modified SCDAP/ RELAP5 on a particular experiment in which a 37-element CANDU bundle was subjected to superheated steam inside a ballooned pressure tube. The focus is on the temperature evolution, the hydrogen accumulation and the metallic melt final relocation geometry.
Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant
Authors: Kawabata Osamu , Ogino Masao
Published: February 2010

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.
Analytical Evaluation of Core Bowing Reactivity in the 4S Reactor
Authors: Nishimura Satoshi , Ohta Hirokazu , Ueda Nobuyuki
Published: February 2010

The 4S (super-safe, small and simple) reactor is a sodium-cooled small fast reactor. The core reactivity is controlled by moving the reflectors installed around the core, and the reactor has a fixed absorber at the core center to accomplish a long core lifetime. To evaluate core bowing behavior and the resulting reactivity feedback in the 4S reactor, an analytical evaluation was conducted under various core power to flow ratios (P/F). The core bowing reactivity under the BOC (beginning of core life) condition becomes increasingly negative with increasing P/F up to 2.0, then becomes less negative with increasing P/F from 2.0 to 3.0, and finally becomes positive at P/F = 3.0. The bowing reactivity under the EOC (end of core life) condition becomes increasingly negative with increasing P/F up to 1.5, then becomes less negative then positive with increasing P/F from 1.5 to 3.0; the core bowing reactivity is positive when P/F ≥ 2.0. These results are mainly caused by the following two mechanisms originating from the structural characteristics of the 4S reactor: - a decrease in neutron absorption by the fixed absorber due to the radial displacement of the inner core subassemblies (under the BOC condition); - a decrease in neutron streaming caused by the small gaps between the outer core subassemblies and the reflectors due to core radial expansion (under the EOC condition).
Assessment of Hill-Side Effect on RV CEDM Nozzles
Authors: Ryu Seung-Bum , Park Moon-kil , Noh Byeong-Wook , Lee Jung-hun
Published: February 2010

CEDM (Control Element Drive Mechanism) controls the power of nuclear power plant by absorbing neutrons produced by nuclear fission. CEDM nozzles are installed in the reactor vessel head by shrinkage fitting and each CEDM nozzle has different length and passes through the spherical head at a different angle which may cause different stress distribution. Formerly, CEDM nozzle at the center line had been considered as a nozzle which is in the most severe condition and also it was selected to get the maximum interaction effects between the discontinuity regions. However, recent reports on nozzles which are installed in a dome infer that there are hill-side effects which may cause higher stresses than stresses on center lined nozzle and those effects should be considered. In this paper, stress distributions at three different locations of the closure head with the CEDM nozzles were evaluated. As a result, the stress intensities at a particular location of the nozzles on the hill-side are higher than those of a nozzle on the center line.
Modeling and Simulation of Multi-Stage Autofrettage in Pressure Vessels
Authors: Malik Afzaal M. , Khushnood Shahab , Gilani Ghulam , Khan Muddasar , Khan Musarrat
Published: February 2010

Thick-walled cylinders play a unique role in cutting edge technologies. These are used in armament industry, aircraft field, automobiles and nuclear reactors. Autofrettage is an effective technique of increasing the fatigue life and bearing capacity of thick-walled cylinders. The most commonly used methods of autofrettage are mechanical, hydraulic and powder gas. In this research paper a study has been carried out of the multi-staging autofrettage process of pressure vessels. The analysis of the autofrettage process is carried out using finite element based software ANSYS, COSMOS. The stress visualization is obtained in the most powerful graphical tool MATLAB. The results obtained are compared with available experimental data.
Protection and Monitoring of Ionizing Radiation - Nuclear Medicine
Authors: Akbar Sajjad , Khalil M. Shahid , Ahmad Shahzad
Published: February 2010

The advancement in technology has resulted into development of Telethrapy and X-ray machine which has high potential hazards of ionizing radiation to user and patient exposed. Ionizing radiations are referred as gamma rays photons. X-rays can cause conjunctivitis and sterility. Ionization radiation is hazard both in radiotherapy and nuclear medicine department. The energy of this radiation is around 10eV, higher the energy of radiation greater is hazard because of penetration into tissues the basic protection rule is either move way from source of radiation or put absorber in between. These equipments are tools of diagnostics, therefore international commission on radiological protection (ICRP) ha recommended that exposure to radiation be kept minimum. Designing of teletherapy facilities play important role in protection and monitoring of radiations. The author has analyzed the protective measures and monitoring of radiations in various hospitals in public and private sector in Rawalpindi / Islamabad Pakistan. It has been observed that only in military hospitals strict protective and monitoring measurers are taken against radiations but in other public and private sector hospitals such measure are compromised due to lack of proper awareness. Pakistan nuclear regulatory authority (PNRA) is taking measures for ensuring protective and monitoring measurer against radiations and arousing awareness to all concerns.
Development of the CAD/MCNP Automatic Conversion Code GEOMIT
Authors: Nasif Hesham R. , Masuda Fukuzo , Iida Hiromasa , Morota Hidetsugu , Sato Satoshi , Konno Chikara
Published: February 2010

GEOMIT is the CAD/MCNP conversion interface code. The old version of GEOMIT had a limited capability from CAD model handling point of view. It is developed to automatically generate Monte Carlo geometrical data from CAD data due to the difference in the representation scheme. GEOMIT is capable of importing different CAD format as well as exporting different CAD format. GEOMIT has a capability to produce solid cells as well as void cells without using complement operator. While loading the CAD shapes (Solids), each shape is assigning material number and density according to its color. Shape fixing process is been applied to cure the errors in the CAD data. Vertices location correctness is evaluated first, then a removal of free edges and removal of small faces processes. Binary Space Portioning (BSP) tree technique is used to automatically split complicated solids into simpler cells to avoid excessive complicated cells for MCNP to run faster. MCNP surfaces are subjected to an automatic reduction before creating the model. CAD data of International Thermonuclear Experimental Reactor (ITER) benchmark model has been converted successfully to MCNP geometrical input. The first wall heat loading calculations agree very well with other countries results.
Preliminary Analysis of Effect of the Intentional Depressurization on Fission Product Behavior During TMLB’ Severe Accident
Authors: Huang Gaofeng , Tong Lili , Cao Xuewu
Published: February 2010

It has been identified that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its effects. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB’ base case and opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomena of revaporization is strong in the RCS.
Implementation of a Pressure Drop Model for the CFD Simulation of Clogged Containment Sump Strainers
Authors: Grahn Alexander , Krepper Eckhard , Weiß Frank-Peter , Alt So¨ren , Ka¨stner Wolfgang , Kratzsch Alexander , Hampel Rainer
Published: February 2010

The present study aims at modelling the pressure drop of flows across growing cakes of compressible, fibrous materials which may form on the upstream side of containment sump strainers after a loss-of-coolant accident (LOCA). The model developed is based on the coupled solution of a differential equation for the change of the pressure drop in terms of superficial liquid velocity and local porosity of the fibre cake and a material equation that accounts for the compaction pressure dependent cake porosity. Details of its implementation into a general-purpose three-dimensional computational fluid dynamics code (CFD) are given. An extension to this basic model is presented, which simulates the time dependent clogging of the fibre cake due to capturing of suspended particles as they pass trough the cake. The extended model relies on empirical relations which model the change of pressure drop and removal efficiency in terms of particle deposit in the fibre cake.
Successive Accident Frequency Assessment From Initiating Event to Containment Failure
Authors: Hirokawa Naoki , Yoneyama Mitsuru , Sato Chikahiro
Published: February 2010

In general, in evaluating large early release frequency (LERF) or containment failure frequency (CFF), level 2 PSA is separated from level 1PSA as described in NUREG-1150 [1]. A brief procedure of this evaluation is that: (1) the core damage frequencies (CDF) to be evaluated for level 1 PSA are grouped into plant damage states (PDSs) based on their accident progression attributes (similar timings to core damage and systems unavailability), (2) frequencies of PDSs are applied to the initial events of event trees (ETs) after the core damage (containment event trees (CETs)), and (3) CETs are developed and quantified. The above method rationally decreases the number of accident sequences to be evaluated in level 2 PSA. In this paper, CETs of internal events for operating state are connected to the end states of the core damage and performed a successive uncertainty analysis from initiating events to containment failures with taking into account for correlations between component failures for a BWR plant (BWR-5 / Mark-II). The rational and some conservative CFF evaluation method are also proposed because the calculation time and computer memory of the above detailed method was very heavy.
Evaluation of Monju Core Damage Risk With Change of AOT Using Probabilistic Method
Authors: Sotsu Masutake , Kurisaka Kenichi
Published: February 2010

MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops which can supply 280 MW of electricity. Limiting conditions of operation (LCO) defined in the safety regulations in MONJU given the allowed outage time (AOT) are evaluated using a PSA technique. The result indicates the possibility of limit extension and some prospects that we should examine.
Evaluation on the Behavior of Stacked Blocks Subject to a Harmonic Excitation
Authors: Choi Woo-Seok , Kim Dong-Ok , Park Keun-Bae , Noh Jae-Man , Lee Won-Jae
Published: February 2010

The core of a VHTR is a stacked structure of graphite blocks. For the development of a seismic analysis methodology, a preliminary evaluation of one single block and single column stacked block structure under a harmonic excitation was performed. In a real design, the blocks are connected by pin and socket component. So, the evaluations were conducted for a stacked structure with and without pin and socket component, respectively. And the evaluation results were compared to each other. For these evaluations, the commercial finite element analysis software ABAQUS/Explicit was utilized.
Network Computing Infrastructure to Share Tools and Data in GNEP
Authors: Kim Guehee , Suzuki Yoshio , Teshima Naoya
Published: February 2010

Network computing infrastructure for sharing tools and data was implemented to support international collaboration. In designing the system, we focused on three issues: accessibility, security, and usability. In the implementation, we integrated existing network and web technologies into the infrastructure by introducing the authentication gateway. For the first issue, SSL-VPN (Security Socket Layer – Virtual Private Network) technology was adopted to access computing resources beyond firewalls. For the second issue, PKI (Public Key Infrastructure)-based authentication mechanism was used for access control. Shared key based file encryption was also used to protect against information leakage. The introduction of the authentication gateway enables to strengthen the security. To provide high usability, WebDAV (Web-based Distributed Authoring and Versioning) was used to provide users with a function to manipulate distributed files through a windows-like GUI (Graphical User Interface). These functions were integrated into a Grid infrastructure called AEGIS (Atomic Energy Grid InfraStructure). Web applications were developed on the infrastructure for dynamic community creation and information sharing. In this paper, we discuss design issues of the system and report the implementation of a prototype applied to share information for the international project GNEP (Global Nuclear Energy Partnership).