bluuebunny/crossref_metadata_2025 huggingface dataset:
Narrative roles in criminal action: An integrative framework for differentiating offenders
Authors: Youngs Donna , Canter David V.
Published: April 2011

Purpose. The proposal that offenders’ narratives help to shape criminal action raises the possibility of a finite set of narrative themes for distinguishing offenders. The present paper seeks to articulate narrative themes that may be active within the roles offenders adopt during offending events. Possible themes may be derived from studies of fundamental narratives in literary criticism, notably Frye (1957). Within personality psychology, McAdams (1993) has also argued for a restricted set of personal narratives in any given culture produced by the dominant dimensions of Potency and Intimacy.

Methods. The sub‐set of narratives on which offenders’ draw can be explored through the roles criminals see themselves as playing during offending episodes. Case study interviews were therefore content analysed to illustrate offence roles based in narrative themes and their cognitive, affective, and identity components.

Results. These considerations suggest that four thematic narrative roles can be distinguished: Professional, Victim, Tragic Hero, and a Revengeful Mission. Distinct patterns of cognitive distortion, affective, and identity components are proposed within these four narrative roles.

Conclusion. The Narrative Offence Roles specified and illustrated in the present paper offer hypotheses for empirical study and the possibility of a new aetiological perspective in criminology.
Contractile and electrophysiological properties of pulmonary artery smooth muscle are not altered in TASK‐1 knockout mice
Authors: Manoury Boris , Lamalle Caroline , Oliveira Roberta , Reid Joy , Gurney Alison M.
Published: April 2011

Non technical summary  Contraction of the muscle in the walls of blood vessels is controlled by the movement of ions into or out of the muscle cells, through proteins called ion channels. By allowing potassium (K+) to leave the cell, K+‐selective ion channels keep arteries dilated. In the pulmonary artery, which carries blood to the lung for gas exchange, a type of K+ channel called TASK‐1 is thought to play an important role in this regulation, but we observed that the pulmonary arteries from mice that lack the TASK‐1 protein do not display any altered properties compared to wild‐type (normal) mice. We found, however, that the pattern of K+ flux in muscle cells from mouse pulmonary artery is different from that of other species, including human. Our results question the use of mouse as a model for human pulmonary artery.
The Electromechanical Low-Power Active Suspension: Modeling, Control, and Prototype Testing
Authors: Evers Willem-Jan , Teerhuis Arjan , van der Knaap Albert , Besselink Igo , Nijmeijer Henk
Published: April 2011

The high energy consumption of market-ready active suspension systems is the limiting factor in the competition with semi-active devices. The variable geometry active suspension is an alternative with a significantly lower power consumption. However, previous designs suffer from packaging problems, nonlinear stiffness characteristics, and failsafe issues. This paper discusses the feasibility of a recently presented, new design, variable geometry actuator, which has a fixed spring and constant stiffness. An actuator model is derived that includes the electric motor and friction characteristics. Using this model, a cascaded controller is developed and the steady-state and dynamic properties are evaluated. The simulation results are validated with prototype tests. The results show a good correspondence between simulations and measurements. Furthermore, a 10 Hz bandwidth can be easily obtained. It is concluded that the electromechanical low-power active suspension design is feasible and that the model gives a fairly accurate representation of both the steady-state and dynamic characteristics of the prototype.
Top Fuel Injection Design in an Entrained-Flow Coal Gasifier Guided by Numerical Simulations
Authors: Wang Ting , Silaen Armin K. , Hsu Heng-Wen , Shen Cheng-Hsien
Published: April 2011

A computational fluid dynamics scheme is employed to simulate the effects of potential fuel injection techniques on gasification performance. The objective is to help design the top-loaded fuel injection arrangement for an entrained-flow gasifier using coal water slurry as the input feedstock. Two specific arrangements are investigated: (a) coaxial dual jet impingement with slurry coal in the center and oxygen in the outer jet and (b) four jet impingement with two single slurry coal jets and two single oxygen jets. When the heterogeneous finite-rate solid-gas reaction scheme is implemented, it is discovered that the particle collision model cannot be implemented with the heterogeneous gasification scheme in the present computational model. The instantaneous gasification model is later employed to examine the particle collision phenomenon by implementing the particle collision model, in which the coal (consisting of carbon and volatiles) is injected as gas, and the water is injected as droplets. The result of droplet tracks shows that the droplets are not bounced around, as speculated, at the intersection where the jets meet, and majority of the droplets pass through the jet impingement section and hit the wall as in the finite-rate case. This implies that the results of the finite rate are acceptable even though the particle collision model is not implemented. The finite-rate result actually presents a worst-case scenario for predicting wall erosion. The particle tracks for both the two concentric and four separate injection configurations show that the coal particles hit the wall and can accelerate the deterioration of the refractory bricks. The case employing two concentric injections provides better fuel-oxidant mixing and higher heating values than the case using four separate injections.
Synthesis of Spatial RPRP Closed Linkages for a Given Screw System
Authors: Perez-Gracia Alba
Published: April 2011

The dimensional synthesis of spatial chains for a prescribed set of positions can be applied to the design of parallel robots by joining the solutions of each serial chain at the end-effector. This design method does not provide with the knowledge about the trajectory between task positions and, in some cases, may yield a system with negative mobility. These problems can be avoided for some overconstrained but movable linkages if the finite-screw system associated with the motion of the linkage is known. The finite-screw system defining the motion of the robot is generated by a set of screws, which can be related to the set of finite task positions traditionally used in the synthesis theory. The interest of this paper lies in presenting a method to define the whole workspace of the linkage as the input task for the exact dimensional synthesis problem. This method is applied to the spatial RPRP closed linkage, for which one solution exists.
Optimizing SCARA Design for Higher Repeatability
Authors: Brethé Jean-François
Published: April 2011

In this paper, the author proposes a new method to design SCARA robots for higher repeatability. First, the author outlines various procedures used in optimal robot design and then points out among the various performance indices those related to repeatability. The author adds some new criteria issued from the stochastic ellipsoid theory. Another innovative part of the paper is to take into account a task-oriented strategy during the design stage, meaning the possibility of adapting task orientation and location in the robot workspace. These concepts are applied to SCARA optimal design. The method described here consists of considering simultaneously robot geometry and joint repeatability, keeping both the reach and the total cost of the sensors constant. It results in an optimization problem with adimensional ratios, which then allows easy comparisons with existing SCARA. The results are surprising and give some clues to answer the underlying question: Are industrial SCARA designed for high repeatability?
Application of Formohydroxamic Acid in Nuclear Processing: Synthesis and Complexation With Technetium-99
Authors: Paviet-Hartmann Patricia , Wright Amber , Mausolf Edward , Campbell Keri , Poineau Frederic
Published: April 2011

Acetohydroxamic acid (AHA) is an organic ligand planned for use in the Uranium Extraction (UREX) process. It reduces neptunium and plutonium, and the resultant hydrophilic complexes are separated from uranium by extraction with tributyl phosphate (TBP) in a hydrocarbon diluent. AHA undergoes hydrolysis to acetic acid which will impede the recycling of nitric acid. During recent discussions of the UREX process, it has been proposed to replace AHA by formohydroxamic acid (FHA). FHA will undergo hydrolysis to formic acid which is volatile, thus allowing the recycling of nitric acid. The reported reduction potentials of AHA and pertechnetate (TcO4−) indicated that it may be possible for AHA to reduce technetium, altering its fate in the fuel cycle. At UNLV, it has been demonstrated that TcO4− undergoes reductive nitrosylation by AHA under a variety of conditions. The resulting divalent technetium is complexed by AHA to form the pseudo-octahedral trans-aquonitrosyl (diacetohydroxamic)-technetium(II) complex ([TcII(NO)(AHA)2H2O]+). In this paper, we are reporting the synthesis of FHA and its complex formation with technetium along with the characterization of FHA crystals achieved by NMR and IR spectroscopy. Two experiments were conducted to investigate the complexation of FHA with Tc and the results were compared with previous data on AHA. The first experiment involved the elution of Tc from a Reillex HP anion exchange resin, and the second one monitored the complexation of technetium with FHA by UV-visible spectrophotometry.
The Application of 10B Enriched Boric Acid in Nuclear Power Industry
Authors: Xu Jiao , Zhang Weijiang
Published: April 2011

Nowadays, boron acid is used widely in nuclear power station as the chemical controlling remedy by making use of 10B as the effective neutron absorbent. But the natural element boron includes 10B just 20% (mass fraction), all the rest is 11B which is very weak at absorbing neutron. This paper brings forward a way to improve the effect of chemical controlling remedy on the nuclear power station through taking 10B enriched boron acid as the substitution of the natural one. If this technological measurement is taken, the operation control of nuclear reaction would get at high level. Not only the amount of boron acid waste would drop suddenly, but the amount of all kinds of chemical agents added in the cooling circulation would lower down sharply. So, many complex chemistry problems in the cooling system could be solved, such as erosion of devices, worse quality of wastewater and consume of chemical agents, etc.
Heat Transfer Analysis in Mixed Convection
Authors: Huang Shan-Fang , Ma Tai-Yi , Gu Han-Yang , Yang Yan-Hua , Yan Xiao
Published: April 2011

Heat transfer is analyzed from a different view in mixed convection in this paper. A concept, namely averaged heat transfer resistance coefficient, is used to describe heat transfer performance. For local position, heat transfer defined by generalized Fourier law is determined by fluid conductance and turbulence heat transfer. On the other hand, heat resistance over the cross section is the integer of the local resistance, where the weight, a function of spatial position, can be expressed by product of local heat transfer and temperature. To enhance heat transfer, it is crucial to reduce the heat resistance where the weight is big, namely near the wall. Heat transfer performance under different buoyancy effect is analyzed by the new. The results show that flow structure and heat transfer are closely connected by a straightforward expression. Heat transfer mechanism of enhancement and deterioration under different stages can be perfectly explained, which can predict heat transfer qualitatively.
Density Functional Calculation of Property of the (1 0 0) Surface of γ-Pu
Authors: Li Rusong , He Bin , Zhang Quanhu , Du Qianwei
Published: April 2011

Pu material can generate surface corrosion and self-radiation effect during storage, leading to the creation and recoil of uranium and helium ions, which produce defects through displacement cascades, these self-irradiation defects tend to change plutonium properties. To study these aging behavior, calculations at the spin unrestricted generalized gradient approximation (GGA) level of density functional theory (DFT) have been performed using the DMol3 programs. Relativistic effects, such as mass-velocity, Darwin term, are considered in this code. Some conclusions are draw as follows: 1) Band structure of the (100) surface of γ-Pu is very narrow around the Fermi level, showing that the eigenstate of this level is mainly composed of local atomic orbital, the local property of electrons in this band is very strong, while the band around the Fermi level is mainly constituted by 5f narrow band; 2) DOS of the (100) surface of γ-Pu are mainly composed of the density of states in −48–41eV, −23–16eV, −3–2eV; 3) Contribution of s shell to the total DOS is mainly distributed in the first interval, and p shell is mainly in the second interval, while d and f shells are mainly in the third interval.
Uncertainty Analysis in Nuclear Criticality Safety Computations
Authors: Ao Qi
Published: April 2011

Uncertainty analysis is imperative for criticality risk assessments when using computational methods to predict the multiplication factor (keff) for fissionable material systems. For the validation of criticality safety computer codes, code accuracy and precision are determined by the computational bias and uncertainty in the bias to account for experimental, computational and model uncertainties. For the application of criticality safety computer codes in the criticality safety design of fissionable material systems, a minimum margin of subcriticality (MMS) must be included to provide additional assurance of subcriticality for any unquantified or unknown uncertainties [1]. Because of a substantial impact of the MMS on nuclear fuel cycle operations, recently increasing interests in reducing the MMS make the uncertainty analysis in criticality safety computations more risk-significant. This paper provides an overview of two most popular keff uncertainty analysis methods for Monte Carlo neutron transport based criticality safety computations: (1) sampling-based methods, and (2) analytical methods. Examples are given to demonstrate their usage for the code validation against benchmark experiments and for the criticality safety design evaluation.
Gamma Spectrum Analysis of Chang’E-1 for Lunar Detection
Authors: Xu HongKun , Fang Fang , Ni Shijun , He Jianfeng , You Lei
Published: April 2011

Gamma-ray spectrum analysis was essential for detecting the elemental abundance and distribution in lunar science. However, for the low-energy region of gamma-ray spectrum, weak peaks were implicated in the fast-decreasing background, and it was difficult to extract characteristic information from original spectra. In order to get a better analytic result, based on wavelet and FFT filtering methods in frequency domain, we had processed the gamma-ray spectrometer (GRS) data of Chang’E-1 (CE-1), and well extracted some useful information of spectral characteristic peaks. Then we preliminarily mapped the distribution of net peak counts for potassium on lunar surface, which indirectly reflected the distribution of elemental abundance. At last, we compared our analytic result with that of Apollo and Lunar Prospector (LP), and found some consistencies and differences.
Experimental Investigation of Heat Transfer in Low Frequency Pulsating Turbulent Flow in Circular Pipe
Authors: Wang Chang , Gao Puzhen , Xu Chao
Published: April 2011

Heat transfer characteristic of low frequency pulsating turbulent water flow in a vertical circular pipe which is heated at uniform heat flux, are experimentally studied under different conditions of Reynolds number, pulsation frequency and relative amplitude. The experiments are performed with the Reynolds number range of 3000 to 20000, pulsation frequency range of 0.033 to 0.1 Hz, and the relative amplitude range of 0.1 to 0.8. This pulsating flow situation is used to simulate the phenomenon happened in the ship power system which is induced by ocean conditions. The effects of pulsation on heat transfer characteristics are presented in terms of relative local and mean Nusselt numbers defined as the ratio of the local and mean Nusselt numbers for pulsation flow to that of the ordinary steady turbulent flow with the same time-averaged Reynolds number Reta. The results show that the relative local Nusselt number is strongly affected by Reynolds number, pulsation frequency and relative amplitude. The phenomena that the Nusselt number would increase or decrease with the increase of the Reynolds number are both observed and the variation is more notable in the entrance region than that in the fully developed region. The relative mean Nusselt number decreases initially as the Reta increases, and then recovers gradually, but finally it has the tendency to decrease again. With the increase of pulsation relative amplitude, the relative mean Nusselt number increases at first and then decreases. And for the Reynolds number range of 3176 to 6670, heat transfer enhancement is observed as the pulsation frequency raises, but complete contrary phenomena appears at Reynolds number range of 11904 to 15844. The obtained heat transfer results are analyzed and seem to be qualitatively in accordance with previous investigations.
A Study of the Effect of Mixed Cores on the Stability of BWRs
Authors: March-Leuba Jose , Wang Weidong , Huang Tai L.
Published: April 2011

Cores loaded with a mixture of fuel types are known to reduce stability margins. Mixed fuel cores have become more common as utilities change fuel suppliers, or when fuel vendors upgrade their fuel designs to take advantage of improved thermal and mechanical margins. This paper studies some of the physical processes that reduce the stability of mixed cores. A number of runs have been performed using the LAPUR6 stability code to evaluate the effect on mixed cores on the stability of a typical BWR. To this end, two fuel types have been set up with two different single-phase to two-phase pressure drop ratios by artificially adjusting the spacer and inlet orifice friction coefficients. The flow and pressure drop characteristics of both fuels have been matched at full flow, full power conditions. All manufacturers match the pressure drop of new fuels so that the flow distributions among the new and old fuel elements operating at the same power are approximately constant. The critical power ratio and thermo-mechanical criteria are typically limiting at full power; therefore matching the flow performance at full power maximizes the margin to these criteria. Stability is of concern at low flows, especially at natural circulation, where the thermal-hydraulic conditions are significantly different from full flow and power. Our simulations show that even if two fuel elements are perfectly matched at full flow, the axial void fraction distribution changes significantly when the flow is reduced to natural circulation conditions and the two fuel elements are not fully thermal-hydraulically compatible at the reduced flows. Basically, the two fuel types set up two separate natural circulation lines, and one of the fuel types essentially starves the other from flow. Since stability has such a strong dependence with channel flow, the reactor stability is controlled by the fuel type that has the smaller flow at natural circulation. A counterintuitive result of this study shows that, in general, loading a more stable fuel type into a mixed core has the opposite effect, and the stability margin of that mixed core is lower until the new, more stable fuel becomes dominant. Because of the burnable Gadolinium in most modern BWR fuels, the highest reactivity fuel elements are the once-burned. Loading a more stable fuel type starves the flow of the high-reactivity older fuel, reducing the stability margin.
Peculiarities of Friction and Depreciation of Contact Surfaces in High-Temperature Lead and Lead-Bismuth Coolants of Reactor Loops
Authors: Antonenkov M. A. , Bokov P. A. , Kuznetsov D. V. , Bokova T. A. , Besnosov A. V.
Published: April 2011

The increasing interest to reactors of heavy nuclear fission by fast neutrons initiates the research substantiating the operational integrity of mechanisms with contact surfaces in a medium of high-temperature heavy liquid-metal coolants.
Research of Radon Transportation Model to Investigate Buried Fault
Authors: Wang Lei , Tuo Xianguo , Liu Chunlai
Published: April 2011

How to get the position, direction and width of buried fault effectively is still very difficult when finding buried faults by measuring 222Rn radioactivity. In this paper we established a technique to carry out buried fault investigation. It was based on Fick’s fist law, Darcy’s law and theory of clusters to analyze radon transportation and simulate 222Rn transportation in ideal conditions. The feasibility indicates that measuring or investigating the concentration of radon to find abnormal region can help people find buried faults. 218Po and 214Po, daughter products of 222Rn, are generally considered to be proportional to initial concentration of 222Rn. 218Po and 214Po have short half-life of 3.05 min and 164us respectively which is very suitable for actual measure work. So in order to accumulate alpha particles effectively, soil gas sampling period is set about twice half-life of 218Po. The established model is applied to analyze two buried fault areas in Southwest China and the results are really much better.
Centrifugal Pump Leak-Free Technology Introduction and Its Applications in EPR Nuclear Power Plant
Authors: Cong Guohui , Zhang Ling
Published: April 2011

Environmental protection requirement is more and more critical now, and it increases the request to prevent dangerous liquid to leak outside in nuclear power plant too. Centrifugal pumps are the most important active equipments in nuclear power plant, but there is a shaft clearance between rotor and stator of centrifugal pump. The shaft clearance can lead pumped fluid to the outside, so the environment may be polluted by the leakage. In some critical conditions such as transferring high radioactive fluid in the pump, the leakage shall be totally forbidden. So solutions have to be found to make centrifugal pumps totally leak-free for applications in nuclear power plant. Normally there are three leak-free technologies for centrifugal pumps: mechanical seal with auxiliary system, canned motor and magnetic drive. In this paper, all the three leak-free technologies and some of their applications in EPR 3rd generation PWR nuclear power plants are presented and discussed. The results show that in EPR nuclear power plant, canned motor pumps can be preferably used for strict environmental requirement of leak-free if the pump power and operating conditions are applicable. For other conditions, pumps with double mechanical seal can also be used with additional sealing water system support. For centrifugal pumps with magnetic drive are not so applicable in high pressure condition, and the safety aspect is weaker than canned motor pumps, generally they are not used in EPR nuclear power plant at present.
Applicability Study of Computed Radiography to NDE for Piping and Component in Nuclear Power Plant
Authors: Hasebe Takashi , Tamai Nobuki , Tatsuno Syohei , Itahashi Yuma , Tokunou Kiyoaki
Published: April 2011

In Radiography Testing (RT) that is an important nondestructive testing as the quality control, the digital imaging technology rapidly provides advancement. The digital imaging technology is more effective than the film method on an environmental side and the cost side, such as unnecessary of the film storage and a chemical treatment by digital output. Especially, in the medical field, the advancement by the digitalized image data processing is remarkable, and it is attempted the upgrade of the inspection technology. However, RT that uses the high-energy radiation and the fast film is a main current in an industrial field, and it has not arrived at digitalization yet. Therefore, in an industrial RT, digitalization is expected and the examination standardization is required also in ASME, JSME, and JIS. We, Mitsubishi Heavy Industries (MHI), studied an applicability of Computed Radiography (CR) to Non-Destructive Examination (NDE) for welds of piping and thick wall component in nuclear power plant. At first, MHI researched image quality of CR for piping. In this research, it was confirmed that the images of testing results by CR method are equivalent to that by film method in terms of visibility of IQI (Image Quality Indicator) and detection performance of welding defects. And we founded the optimized shooting conditions for piping. Second, MHI researched image quality of CR for thick wall component. In the result, the noise shown in fig.1 was occurred on CR image when the thick wall component such as pressure vessel is radiographed with high energy. It is speculated that the primary cause of this noise is the scattered X-ray effect (shown in fig.2). Therefore it is necessary to investigate the effect of the scattered X-ray on CR image. In this study, to reduce the scattered X-ray effect on CR image, we investigated the effect of 1) screen, 2) screen + filter on image quality of CR for thick wall component. And we studied the optimized shooting conditions and parameters for thicker component than piping to aim for more application. Finally, we applied CR to the pipings and components for nuclear power plant with the optimized shooting conditions and parameters.
Validation of a New Lattice Physics Code GALAXY for PWRs
Authors: Yamaji Kazuya , Koike Hiroki , Sato Daisuke , Tsubota Shinobu , Matsumoto Hideki
Published: April 2011

A new lattice physics and assembly calculation code GALAXY with the 172 energy-group ENDF/B-VII.0 library has been developed. GALAXY generates few group nuclear constants used in a new core simulator COSMO-S. The GALAXY code uses the many enhanced calculation method for more explicit treatment of neutronics characteristics. The outline of enhanced methods used in GALAXY and the qualification results are shown in this paper. From the qualifications in the continuous energy Monte Carlo benchmark, critical experiment analyses and post irradiation examination (PIE) analyses, GALAXY with the library was validated and the applicability of GALAXY to PWR nuclear design was confirmed.
The Improvement Research of Accident Generation Mechanism of Dynamic Event Tree
Authors: Zhang Yongfa , Tong Jiejuan , Zhou Yu
Published: April 2011

Accident generation mechanism is a very important part of the dynamic reliability methodologies. CCMT and DET are two well known dynamic methods. The accident generation mechanism of them is base on the fixed time step, this mechanism has two remarkable deficiencies, one is that it unable to accurately consider the occurrence time of demand conversion; another is that if the time step is long, the traditional mechanism unable to treat the possible twice or even more demand conversion. In order to overcome these deficiencies, the paper proposed a new algorithm which evolved both time and processing variable. The basic principle is provided. some common problems are also discussed. At last, the computer execution algorithm is advanced, and a example system is provided to demonstrate the power of new mechanism. The analysis results show that the new mechanism is effective.
Numerical Analysis of Thermal Stratification in Pressurizer Surge Line
Authors: Qi Zhanfei , Cao Xuewu
Published: April 2011

The thermal stratification is generic to un-isolable piping system in pressurized water reactor (PWR), which is not considered in original design. This effect can finally threaten the integrity of the piping system. The pressurizer surge line is affected by thermal stratification during reactor heat up and shutdown processes particularly. It’s important to investigate and mitigate thermal stratification in the surge line to meet safety requirements. In this paper, the thermal hydraulic analysis and the mechanical analysis using numerical simulation method were conducted to evaluate the impact of thermal stratification in the surge line during reactor heat up processes. The results give the three-dimensional temperature distribution and stress status of the surge line. It indicates that thermal stratification can cause stress concentration in the displacement restricts. The mitigation of thermal stratification was also studied. The numerical simulation of pressurizer surge line can put forward thermal stratification analysis.
Analysis of ABWR Critical Control and Heat-Up Control Operation by TRACG Code
Authors: Ishii Yoshihiko , Kitou Kazuaki , Ikegawa Tomohiko , Hasegawa Shin , Ochi Hitoshi
Published: April 2011

Most startup and shutdown operations in advanced boiling water reactors (ABWRs) are automated by an automatic power regulator (APR). Hitachi and Hitachi-GE utilized the three-dimensional transient analysis code TRACG to design and verify the APR control algorithms. To verify the algorithms, an external neutron source model that makes it possible to simulate a sub-critical initial core, a water temperature reactivity model, a startup range neutron monitor (SRNM) model, and the APR system models were developed and coded onto the TRACG code. The improved TRACG code has been tested and verified with ABWR startup test data. In the test, the criticality was achieved 40 min after beginning of control rod (CR) withdrawal. The code results, for example, CR operation timing, CR withdrawal length, and signals of the neutron sensors agreed well with the test data. In the heat-up control mode, the measured increasing rate of the reactor water temperature was well simulated for a period longer than six hours.
Synergistic Effect of Neutron and Gamma Irradiation on CMOS Integrated Circuits
Authors: Liu Yan , Jin Xiaoming , Chen Wei , Lin Dongsheng , Yang Shanchao
Published: April 2011

To investigate the interaction between the total ionizing dose (TID) effect and displacement damage, this paper presents the experiment results of several CMOS integrated circuits’ radiation effect in a mixed neutron and gamma ray environment. The electrical and functional tests were made in four irradiation environments: 1. neutron (high neutron/gamma ratio); 2. gamma ray; 3. combined irradiation of neutron and gamma ray (low neutron/gamma ratio); 4. TID test after neutron irradiation. All the parameters were tested online. Compared with the single radiation environments, results show that the mixed neutron and gamma irradiation can induce a synergistic effect on the samples. In the neutron and gamma synergistic irradiation, the devices exhibit considerably severer degradation. The synergistic effect occurs while neutron and gamma exist synchronously. Where the percentage of gamma is comparable or larger than neutron at the levels of concern, one can expect synergistic effects will occur in the CMOS devices, and the TID effects should not be decoupled from nonionizing or displacement damage.
Simulation of CCD Degradation Induced by Protons Irradiation
Authors: Wang Zujun , Huang Shaoyan , Liu Minbo , Tang Benqi , Xiao Zhigang , Zhang Yong
Published: April 2011

The mechanisms of charge coupled devices (CCD) irradiated by protons are analyzed. The simulation models of ionization damage and displacement damage are developed. The charge transfer efficiency (CTE) decreased by proton irradiation is numerically simulated. The CTE degradation caused by different traps and by protons with different energies has been studied respectively. Both surface dark signals induced by proton ionization damage and bulk dark signals induced by proton displacement damage are numerically simulated. The variability of surface dark signals, bulk dark signals, and total dark signals with proton fluence is compared. The simulation results are in agreement with the experimental results of the relevant literatures.
Numerical Analysis of Safety Injection System’s Flow Rate
Authors: Sun Dengke , Wang Xiaojiang , Li Jun , Liu Jiang , Li Changyue
Published: April 2011

In this paper, a hydraulic model for Safety Injection System (SIS) of M310 reactor is extended. The model is checked and calibrated by test results under test conditions. Based on commissioning test criteria, the system’s maximum and minimum pressure drop coefficients are calibrated according to anti-extrapolation method. Considering modifications of various projects, analysis of 41 flow rate curves under different conditions has been performed using this model. These flow rate curves indicate the relationship between injection flow rates and primary circuit pressures. Accident analysis has taken these curves as input data. Also, the strategy for dealing with accident is established based on these curves. Results of accident analysis show that the design of SIS system can satisfy the safety requirements of M310 reactor. The sensitivity analysis of typical conditions illustrates that injection flow rate will increase as the primary circuit pressure decreases. With the same configuration, the injection flow rate during recirculation phase will be smaller than that during direct injection phase, which is mainly caused by the decrement of suction elevation and the increment of fluid temperature. When low head safety injection pump (LHSI PO) is boosting high head safety injection pump (HHSI PO), if the pressure is relatively high, the injection flow rate will not be improved apparently. If the pressure is relatively low, the boosting is necessary. These conclusions can be the basis for the later optimization design.
Environmental Considerations for Preparing Permit Applications of New Nuclear Power Plants Located in Greenfield Sites
Authors: Wan Ping K. , Chan Desmond W. , Carson Alice C.
Published: April 2011

Nuclear power generation has become an increasingly attractive alternative in the United States (U.S.) power market due to several factors: growing demand for electric power, increasing global competition for fossil fuels, concern over greenhouse gas emissions and their potential impact on climate change, and the desire for energy independence. Assuring the protection of people and the environment are of paramount concern to nuclear power generators and regulators as we move towards a possible nuclear renaissance. Thus, sound engineering design is of utmost important and potential environmental and safety concerns must be carefully evaluated and disposition during permitting of the new nuclear power plants. Areas to be considered in order to alleviate these concerns include the following: • Site meteorology and dispersion conditions of the area; • Evaluation of radiological consequence during normal plant operation and emergency conditions; • Water availability for plant cooling system; • Evaluation of potential land use, water use, ecological and socioeconomic impacts of the proposed action. This paper focuses on site suitability evaluation for greenfield sites through site characterization, examination of challenges/constraints in deployment of available technology/plant systems, and mapping of permitting compliance strategy. Case studies related to selection of plant systems based on the environmental site conditions, preferred compliance plan, and public acceptance, are included.
Determination of Diffusion Parameters by a Computer Code (FDP) Based on Mathematica 6.0
Authors: Wu Tao , Samer Amayri , Drebert Jakob , Van Loon Luc R. , Reich Tobias
Published: April 2011

A computer code named Fitting for Diffusion Parameters (FDP) based on Mathematica 6.0 has been developed for modeling through- and out-diffusion experiments. FDP was used to determine the diffusion coefficients (De) and the rock capacity factors (α) for tritiated water (HTO) and 22Na+ and the distribution coefficient (Kd) of 22Na+ in Opalinus Clay (OPA). The values for De and α were obtained by fitting the results of experimental data of both transient and steady-state phases to the analytical solution of accumulated activity. The quality of the parameters De and α was tested by using them as input parameters in the equation of flux. Moreover, the diffusion parameters of HTO and 22Na+ were determined also by out-diffusion experiments. Under ambient condition at pH 7.6, the De value of (1.5 ± 0.1) × 10−11 m2/s for HTO is lower than that of (1.9 ± 1.1) × 10−11 m2/s for 22Na+, which could be explained by the electrostatic attraction between the negative surface charge of OPA and the sodium cations. For the non-sorbing species HTO, α was 0.15 ± 0.01. For the weakly sorbing species 22Na+, α was 0.50 ± 0.02 and Kd equaled (1.5 ± 0.3) × 10−4 m3/kg. The obtained diffusion parameters for HTO and 22Na+ in OPA are in good agreement with previous results by Van Loon et al. [1, 2]. FDP developed in this study has been used successfully to determine the parameters De and α for the diffusion of 237Np(V) in OPA [3].
Numerical Study on Flow and Heat Transfer in Concentric and Eccentric Annuli
Authors: Yang Xingbo , Su Guanghui , Tian Wenxi , Wang Jiayun , Qiu Suizheng
Published: April 2011

The narrow annular channel has been widely studied for its relatively larger heat transfer surface and structural compatibility. In this study, numerical studies have been performed on the 3D forced flow and heat transfer of water in concentric and eccentric annuli by using CFX codes. The gaps of concentric annuli range from 1.0mm to 4.0mm with the interval of 0.5mm. The eccentricity ratioes in eccentric annuli are 0.2, 0.3, 0.5 and 0.7. The radius ratioes of the eccentric annulus include 0.33, 0.5 and 0.66. The calculated results are compared with some experimental data and they agree well. The results show that the flow frictional resistance factor decreases with increasing the gap size. The impact of gap sizes on the flow frictional resistance factor decreases with increasing Reynolds number. The flow frictional resistance factor in the eccentric annuli is larger than that in concentric annuli. Furthermore, the effects of the eccentricity ratio and gap size on Nu number and the flow frictional resistance factor are also investigated.
Experimental Test Plan of Air Ingress for HTGR
Authors: Terada Atsuhiko , Yan Xing L. , Hino Ryutaro , Sato Hiroyuki
Published: April 2011

Primary pipe rupture is an important design base accident in the high temperature gas cooled reactor (HTGR). When a primary pipe of the HTGR ruptures, helium coolant gas in the reactor blows out into the reactor confinement structure and the reactor primary system depressurizes. After the pressures of the reactor and the confinement equalizes, air is expected to enter the reactor core from the breach. Consequently, the core graphite structures may be oxidized by the air and the complicated natural convection of multi component gas mixtures with chemical reactions would take place inside the reactor. Hence, it is necessary to investigate the air ingress process, the natural convection of multi component gas mixtures in order to understand the effect and develop mitigation of the air ingress. JAEA has performed analysis and fundamental experiments about air ingress from the rupture of one or more main coolant pipes on the lower body of the RPV. These studies showed the air ingress phenomena in the depressurized reactor and proposed a new passive mechanism of sustained counter air diffusion (SCAD) that has been shown effective in preventing major air ingress through natural circulation in the reactor. In the present plan, JAEA will construct an experimental reactor mockup including reactor core, the SCAD system, pressure vessel, coaxial pipe and so on. The core is made of graphite or ceramics and heated by electric heaters to allow for test operation up to 1200°C. Present status of these activities will be presented. Based on the analytical results and know-how obtained through the bench test, a 1/8 scale air ingress mockup test, which intends to simulate the accident condition of GTHTR300, is being planned with a conceptual design as the next step of the air ingress experiment evaluation in JAEA. In the design of mockup experimental facility, it is important to reproduce flow phenomena in a reasonable scale from the viewpoint of construction cost. We designed the internal structure to reproduce mixing performance of multi-component flow involving ingress phenomena especially in the guillotine breaks of primary coolant pipe. Complex flow pattern with gas oxidized chemical interaction in the graphite porous structure of the HTGR core will be characterized. Preliminary analytical results especially natural circulation flow patterns induced by density and concentration difference obtained with a CFD model agreed well with that measured by bench experiments, which showed natural circulation pattern in a simplified reactor.
PKL Tests on Heterogeneous Inherent Boron Dilution Following SB-LOCA: Applicability to Reactor Scale
Authors: Umminger Klaus , Schollenberger Simon Philipp , Cornille Se´bastien , Agnoux Claire , Quintin Delphine , Freydier Philippe
Published: April 2011

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.
Experience on Transient Thermal Modeling of Sodium to Sodium Heat Exchanger Using FDM and FVM
Authors: Agarwal Sourabh , Revathy K. , Banerjee I. , Padma Kumar G. , Babu C. A. , Vaidyanathan G.
Published: April 2011

Several intermediate heat exchanger (IHX) modelling techniques were examined, in order to predict the outlet temperature of primary and secondary sodium at different operating conditions. In the present study, two different approaches namely the Finite Difference Method (FDM) with nodal heat balance and modified nodal heat balance schemes; and Finite Volume Method (FVM) using simple upwind, exponential extrapolation and QUICK schemes have been attempted.
Visual and Thermal Studies of Single Tube Reflood Under Typical PWR LB-LOCA Conditions
Authors: Zeng Yiyun Jessy , Hale Colin P. , Walker Simon P. , Hewitt Geoffrey F.
Published: April 2011

The common approach to safety in a nuclear power plant is to design the system to respond safely to a large postulated accident, the so-called design basis accident. Accidents more severe than the design basis accident (“severe accidents”) are assessed but the system is not designed to withstand them; they are considered too unlikely to require specific design actions. For the pressurised-water reactor (PWR), the design basis accident (DBA) is the Large Break Loss-of-Coolant Accident (LB-LOCA), in which it is assumed that one of the large inlet coolant pipes from the circulating pump to the reactor vessel is completely broken and moves apart to allow free discharge of the primary coolant from both broken ends. For this type of break total coolant loss will occur in 100s or less. Although the reactor is by this time sub-critical so that little power is produced from fission, a large amount of decay power exists and causes the fuel rod claddings to have a temperature in the region of 600–800 °C. This paper describes experimental investigations designed and performed in order to provide detailed information about the macroscopic behaviour of the steam-water flow occurring during the reflood phase following a PWR LB-LOCA. Specifically, a bottom-up rewetting process was studied, in which water droplets may be entrained in the vapour flow and contribute to cooling of the hot fuel pin before it is quenched. In these experiments the test section is initially preheated to temperatures up to 600 °C and then quenched by introducing water at the bottom of the tube at atmospheric pressure. During the course of this transient process axial temperature and heat flux profiles will be recorded, extending the existing databank of cases for code validation. Simultaneously, an axial viewing technique will be applied to observe the quench front, and any pre-cursory droplet production, occurring during these singletube reflood experiments. As part of the preliminary validation of this novel technique, a series of air-water vertical upflow conditions have been examined. The results of these preliminary studies provide detailed visualisation of typical entrainment processes likely to be encountered during single-tube reflood.
Simulation of Fast Gas Injection Expansion Phase Experiments
Authors: Pellini Donella , Maschek Werner , Forgione Nicola , Poli Francesco , Oriolo Francesco
Published: April 2011

The injection of a high pressure gas into a stagnant liquid pool is the characteristic phenomenon during the expansion phase of a hypothetical core disruptive accident in liquid metal cooled fast reactors. In order to investigate lot of mechanism involved in this phase of the accident’s evolution, an experimental campaign called SGI was performed in 1994 in Forschungszentrum Karlsruhe, now KIT. This campaign consists of nine experiments which have been dedicated to assess the effects of different pressure injection, the nozzle’s size and the presence of inner confinement in the formation of the rising bubble. Three of these experiments, which were focused on the pressure effects, have now been simulated with SIMMER III code and with FLUENT 6.3 code, a commercial CFD code. Both codes, despite their different features, have showed a good agreement with the experimental results. In particular, time trend evolutions of pressures and bubble volumes have been well reproduced by simulation. Furthermore, both codes agree on the shape of the bubble, even though they have evidenced same discrepancies with the experimental shape.