bluuebunny/crossref_metadata_2025 huggingface dataset:
RMC1.0: Development of Monte Carlo Code for Reactor Analysis
Authors: She Ding , Xu Qi , Wang Kan , Yu Ganglin
Published: April 2011

This paper describes a newly developed Monte Carlo code used for reactor analysis called RMC1.0, which is based on ACE format library. RMC1.0 is able to estimate criticality eigenvalue, and tally flux/spectrum with collision estimation method or tracking length method. Series of benchmarks and other examples are calculated for validation, which prove that RMC1.0 gives a good performance in both accuracy and efficiency compared with mcnp5. Despite its limitation in geometry processing, RMC1.0 has made a profitable attempt in self-development of Monte Carlo code for reactor analysis.
Apply Performance Evaluation Matrix on Investigating Human Error Events in the Main Control Room
Authors: Yenn Tzu-Chung , Jou Yung-Tsan , Lin Chiuhsiang Joe , Tsai Wan-Shan , Hsieh Tsung-Ling
Published: April 2011

Digitalized nuclear instruments and control systems have become the main stream design for the main control room (MCR) of advanced nuclear power plants (NPPs) nowadays. Digital human-system interface (HSI) could improve human performance and, on the other hand, could reduce operators’ situation awareness as well. It might cause humans making wrong decision during an emergency unintentionally. Besides, digital HSI relies on computers to integrate system information automatically instead of human operation. It has changed the operator’s role from mainly relating operational activity to mainly relating monitoring. However, if operators omit or misjudge the information on the video display units or wide display panel, the error of omission and error of commission may occur. Therefore, how to avoid and prevent human errors has become a very imperative and important issue in the nuclear safety field. This study applies Performance Evaluation Matrix to explore the potential human errors problems of the MCR. The results show that the potential problems which would probably affect to the human performance of the MCR in advanced NPPs are multiple accidents, pressure level, number of operators, and other factors such as working environmental.
Study of Self-Shielding Calculation With Temperature Distribution Applying Wavelets Scaling Function Expansion Continuous-Energy Resonance Method
Authors: Yang Weiyan , Wu Hongchun , Cao Liangzhi
Published: April 2011

The radial power distribution within the fuel rod is important in fuel integrity evaluation. In this paper, the wavelets scaling function expansion method is applied to evaluate the radial power distribution and obtain continuous-energy spectrums in the fuel rod with a temperature distribution. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are processed by NJOY, while the multi-group nuclear data library is jeff31 issued by IAEA. However, for different temperature problems, because of Doppler effect, the continuous-energy data library is different, and should be updated by applying NJOY. The calculation efficiency is a problem needed to be improved. Therefore, in this paper, the interpolation is utilized to obtain the continuous-energy cross-sections for other temperatures between those given temperatures. Also the precision of the temperature interpolation is discussed. Finally, the differences of continuous-energy spectrums, reaction rates and k-inf results are presented and compared with MCNP calculation.
Study on Liquid Lithium Target System for Boron Neutron Capture Therapy (BNCT)
Authors: Takahashi Minoru , Kobayashi Tooru , Nakatsuka Masashi , Ardiansyah Teddy , Kulha´nek Martin , Voja´cˇek Alesˇ , Dosta´l Va´clav , Uchida Shoji , Zhang Mingguang
Published: April 2011

The concept of the liquid lithium target system for the boron neutron capture therapy (BNCT) using an accelerator was formulated. A lithium flow loop was designed and fabricated for development of lithium jet nozzles that provide a stable lithium plane jet. A water flow loop was designed, fabricated and used for development of lithium jet nozzles. Experiment of water plane jet was conducted to investigate the stability of high velocity plane jets. A large amount of droplets were entrained from the jet surface when the jet was unstable. The jet was observed by being illuminated with a stroboscope. It was found that a stable plane jet was realized by using a nozzle with the gap of 0.5mm and the length of 70mm. It was concluded that a fully developed flow without inlet disturbances at the outlet of the nozzle was required for the stability of the high velocity plane jet.
Model Development of Turbulent Dispersion Force for Advanced Two-Fluid Model in Consideration of Bubble-Liquid Phase Interactions
Authors: Hosoi Hideaki , Yoshida Hiroyuki
Published: April 2011

Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase simulation method such as interface tracking method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as rod bundles. However, since two-fluid model include a lot of constitutive equations, applicability of these constitutive equations must be verified by use of experimental results, and the two-fluid model has problems the result of analyses depends on accuracy of constitutive equations. To solve these problems, an advanced two-fluid model has been developed in Japan Atomic Energy Agency. In the model, an interface tracking method is combined with the two-fluid model to predict large interface structure behavior accurately. Liquid clusters and bubbles larger than a computational cell are calculated using the interface tracking method, and those smaller than a cell are simulated by the two-fluid model. Constitutive equations to evaluate the effect of small bubbles or droplets on two-phase flow required in the advanced two-fluid model as same as a conventional two-fluid model. However, dependency of small bubbles and droplets on two-phase flow characteristic is relatively small, and the experimental results to verify the equations are not required much. The turbulent dispersion force term is one of the most important constitutive equations for the advanced two-fluid model. The turbulent dispersion force term has been modeled by many researchers for the conventional two-fluid model. However, the existing models include effects of large bubbles and deformation of bubbles implicitly, these models are not applicable to the advanced two-fluid model. In this study, we develop the new model for turbulent dispersion force term. In this model, effect of large bubbles and deformation of bubbles are neglected. The liquid phase turbulent kinetic energy and bubble-induced turbulent kinetic energy are considered to evaluate driving force in the turbulent diffusion of small bubbles. The bubble-induced turbulent kinetic energy is given by the function of bubble diameter and local relative velocity, and the liquid phase turbulent kinetic energy is similar to the single phase flow case. Furthermore, we considered energy transfer from the bubble-induced kinetic energy to the liquid phase turbulent kinetic energy. To verify the developed model, the advanced two-fluid model and the model for turbulent dispersion term were incorporated to the 3-dimensional two-fluid model code ACE-3D, and comparisons between the results of analyses and air-water two-phase flow experiments in 200 mm diameter vertical pipe were performed.
Effect of Pin and Subassembly Heterogenity in Sodium Void Worth Calculations
Authors: Raghu Kumar G. , Reddy C. P. , Sathyamoorthy V.
Published: April 2011

Metal fuelled sodium cooled fast reactors are known to have high breeding ratio and short doubling time. Due to these features they play a very important role in the energy scenario, where higher power growth is required. Large sodium cooled fast reactors have positive sodium void coefficient, which is considered to be undesirable feature even though reactor safety can be established for all design based accidents like loss of flow and transient over power accidents. These types of fast reactors, which have harder neutron spectra are having higher sodium void coefficient compared to ceramic fuelled fast reactors. In many of the safety analysis the total sodium void is calculated and it is used in the safely evaluation. However the sodium in the metal fuelled reactor has got three parts, namely bonding sodium, coolant sodium and the sodium in the inter space of subassembly hexagonal cans. In the reactor accident scenario the behavior of these three components of sodium will be different and will effect the sequence of the accident. The finer details, of the fuel sub assembly, are modeled in to Monte Carlo code and the sodium void coefficient is calculated for each of the component for the fuel zones. This study will be helpful in improving safety of the reactor and also reducing the conservatism in the safely features.
The Mechanical Properties and Friction and Wear Behavior of C/C-SIC
Authors: Wen Gao , Long Chongsheng , Rui Tang , Wang Jiping
Published: April 2011

Carbon fiber reinforced carbon-silicon carbide composites (C/C-SiC) were prepared by chemical volume infiltration (CVI) method and reaction melt infiltration (RMI) technique of silicon liquid to carbon reinforce carbon matrix composites. The friction and wear behaviors of C/C-SiC composites at various loads and sliding speeds were investigated by MRH-3 block-on-ring tribometer at room temperature under water lubricating conditions. Furthermore, the morphologies, phase of the worn surface and the debris were observed, examined and analyzed by scanning electron microscopy (SEM) and energy-dispersive X-ray microanalysis (EDAX) respectively. Experimental results showed that the C/C-SiC composites had a better wear resistence, and the friction coefficient under water lubricated conditions is about 0.02–0.06. The influence of sliding speed on the friction coefficients and the specific wear rate of C/C-SiC is more obvious than that of normal load when the load is less than 200N (inclueded200N). The friction coefficient and the specific wear rate of C/C-SiC decreased as the sliding velocity increased. At the sliding speed higher than 2m/s, the friction coefficient is less than 0.02. The specific wear rates is at a low level about (2×10−7mm3/Nm–5×10−8mm3/Nm).
Experimental Investigation of Freezing Behavior of Melt/Solid Particles Mixture on to Metal Structure
Authors: Soner Md. Abdul Malek , Seo Shunichi , Hasegawa Yu , Himuro Yusuke , Morita Koji , Matsumoto Tatsuya , Fukuda Kenji , Maschek Werner
Published: April 2011

The freezing and penetration of molten core fuel and structural materials penetrating into flow channels are important thermal-hydraulics phenomena to safety assessment of postulated core disruptive accidents in liquid metal reactors. The main objective of this study is to investigate fundamental characteristics of freezing and penetration behavior involved in melt and solid mixture flowing on-to structure material. In our study, solid copper particles mixed with molten wood’s metal (melting point 78.8°C) was used as a simulant melt, while stainless steel and brass were used as freezing structures. A series of fundamental experiments was performed to study the effects of solid particles on the freezing and penetration behavior under the various thermal conditions of molten metal and varying solid particle volume fraction and structure metal. The melt flow and distribution were observed using a digital video camera. The melt penetration length on the structure and proportion of adhered frozen metal on to structure surfaces were measured in the present series of experiments. The results indicate that penetration length becomes shorter with increasing solid particles volume fraction in melt. The present results will be utilized to build a relevant database for verification of fast reactor safety analysis codes.
Designing DOE’s NGNP: Don’t Forget High-Temperature Gas Reactor (HTGR) Experience
Authors: August J. K. , Hunter J. J.
Published: April 2011

The U.S Department of Energy’s (DOE) Next Generation Nuclear Plant (NGNP) has three goals in its mission: deliver (1) electricity, (2) process steam and heat, and (3) generate hydrogen to support a hydrogen economy. DOE selected the high-temperature, helium-cooled graphite (moderated) reactor (HTGR) technology for its fundamental design. However, the NGNP faces many challenging design, licensing and cost hurdles. With huge risks and its associated expenses, the NGNP — a completely new reactor technology — requires an open-minded understanding, consideration and application of earlier HTGR design features. This paper suggests reconsidering some HTGR features in an updated Generation IV HTGR as a design alternative.
Modeling of NPP Canned Motor Pump Performance Degradation Assessment Based on Multi-Agent System
Authors: Zhang Liming
Published: April 2011

The functional hierarchy and the performance degradation characteristic of the canned motor pump (CMP) used in nuclear power plants (NPP) are analyzed. And to model the performance degradation of CMP, a hierarchical model based on multi-Agent system (MAS) is proposed. In the model, the framework of the system is designed and the disadvantages of current assessment are conquered. According to the function structure of CMP, the performance degradation and prediction Agents for each hierarchy are divided and the concrete functions for each Agent are also defined. The MAS model provides a guidance template to implement equipment performance degradation and prediction system.
The Analytical and Experimental Research of In-Vessel Corium Retention in CPR1000
Authors: Chen Xing , Zhang Shishun , Lin Jiming , Zhang Huiyong
Published: April 2011

The analytical and experiment research of In-Vessel Corium Retention (IVR) in the Chinese Pressurized-water Reactor 1000 MWe (CPR1000) are introduced. The IVR research consists of preliminary phase and detailed phase. The analysis of thermal failure, structural failure and penetration failure of Reactor Pressure Vessel (RPV) and the experimental research of External Reactor Vessel Cooling (ERVC) are performed at preliminary phase. Analysis results show that the RPV failure is the dominated by thermal failure mode and the probability of the thermal failure is very low. Test results show that the IVR success probability for CPR1000 is about 99% if the Critical Heat Flux (CHF) of CPR1000 is the same as that of AP600. Further works, including the ERVC enhancement design, the CHF test of the RPV outer wall and the recalculation of the IVR success probability for CPR1000, will be performed at detailed phase in the near future.
The RELAP5 Simulation of 2″ LOCA of APEX Facility
Authors: Xie Heng
Published: April 2011

The RELAP5/SCDAP Mod3.2(am5) code is employed to simulate the OSU-AP1000-05 test conducted in the Advanced Plant Experimental (APEX) test facility at Oregon State University (OSU). The APEX-1000 test facility is an one-fourth height, one-half time scale, and reduced pressure integral systems facility to simulate the Westinghouse Advanced Passive 1000 MW (AP1000) pressurized water reactor. OSU-AP1000-05 is a two-inch break at the bottom of cold leg #4 with 3 out of 4 ADS-4 valves of OSU-APEX-1000 facility. RELAP5 predictions are compared to the experimental data generated by the test. The comparison shows good agreement between the predicted and measured sequence of events of some key parameters during the transient. From the comparison results, it could be preliminary concluded that the RELAP5/SCDAP Mod3.2(am5) code are suitable to simulate the small LOCA of APEX.
Simulation of a Steam Generator Tube Rupture Accident in a Lead-Cooled Accelerator Driven System
Authors: Flad Michael , Wang Shisheng , Maschek Werner
Published: April 2011

The European Facility for Industrial Transmutation (EFIT) is developed to transmute long-lived actinides from spent fuel on an industrial scale. In this lead-cooled reactor an intermediate loop is eliminated for economic reasons. Within the framework of design and safety studies the impact of a steam generator tube rupture accident has been investigated. In this postulated event high-pressured liquid water blasts into the lead pool which could trigger various transients. As a major concern steam could be dragged into the core featuring a positive void worth. A thermal lead/water interaction could lead to in-core damage propagation; it could initiate a sloshing of the lead coolant and trigger voiding processes. Furthermore the pressurization of the cover gas needs to be considered. To prove the feasibility of the proposed design these risks are investigated and assessed. Numerical simulations are performed using the advanced safety analysis code SIMMER-III [2]. For the important issue of thermal lead/water interactions the SIMMER code has been validated against Japanese heavy-liquid/water injection experiments.
Some Nuclear Reactor Safety Related Aspects of Plunging Jets
Authors: Krepper Eckhard , Weiss Frank-Peter , Alt So¨ren , Kratzsch Alexander , Renger Stefan , Ka¨stner Wolfgang
Published: April 2011

Plunging jets play an important role in nuclear reactor safety research. In the present paper the case of the strainer clogging issue is considered. Entrained air caused by a plunging jet has an influence of the liquid flow field and on the fibre transport in the sump. In the paper the amount of entrained air is given as an inlet boundary condition according to correlations in the literature and confirmed by own experiments. The influence of entrained air on the fibre deposition pattern at the bottom of a tank and on the mixing procedure for the case of temperature differences between jet and tank water are investigated by CFD calculations and compared to experiments. The presented work is part of a joint research project performed in cooperation between the University of Applied Science Zittau/Go¨rlitz and Forschungszentrum Dresden-Rossendorf. The project deals with the experimental investigation of particle transport phenomena in coolant flow in Zittau and the development of CFD models for its simulation in Rossendorf (Krepper et al. 2008).
Steady-State and Transient Analysis in Single-Phase Natural Circulation of ABV-6M
Authors: Wang Fangnian , Tian Zhaofei , Liu Jiange
Published: April 2011

This paper discusses the results of the thermal-hydraulic investigations of natural circulation. The RELAP5/MOD3.4 computer code has been used to simulate the natural circulation in an ABV-6M Nuclear Power Plant (NPP), and steady-state and transient operational characteristics are researched. Some results can be listed as follows. In the steady-state analysis, the relationship between the reactor power and the mass flow rate of primary loop, and the temperature difference of core inlet and outlet are obtained. The comparisons between the RELAP5 results and the theoretic calculation data indicate good general agreement. Besides, the mass flow rate of natural circulation will increase and the flow will become stable more quickly when the reactor power increases on the condition of the system with a fixed construction. The trends of the thermal parameters are obtained and the temperature envelope curve of secondary circuit fluid is ascertained. In the transient analysis, some perturbations, such as the fluctuations of feedwater temperature and feedwater mass flow rate, have influence on the system performance. Then, the comparisons between taking an action and not taking an action on Nitrogen Pressurizer are obtained when these small perturbations occur. It is turn out that the safety of natural circulation loop is sensitive to some small perturbations.
A Novel Radioisotope Microbattery Based on Betavoltaic Effect and Work Function
Authors: Chen Haiyang , Lan Jiang , Zuo Fuchang , Sun Min
Published: April 2011

This paper reports a novel radioisotope microbattery structure that integrates a betavoltaic converter with a work function converter. The battery collects energy from radioisotope and environment vibration. A model is developed to simulate the mechanism of the proposed battery, and select parameters to improve its efficiency. Using the proposed model, the battery is designed with structures optimized for the environment vibration frequency in the range of 100–400Hz and 63Ni of 11mCi. The theoretical output power is on the order of 200nW. The output power collected from the radioisotope is close to that from environment vibration. Since the vibration beam frequency of the work function converter is much larger than the environment frequency, the output power of the battery keeps stable when the environment frequency changes significantly.
Study on Countercurrent Flow in a PWR Hot Leg Under Reflux Cooling
Authors: Minami Noritoshi , Murase Michio , Tomiyama Akio
Published: April 2011

In this paper, results of experiments and numerical simulations for counter-current flow in a pressurized water reactor hot leg under reflux cooling are summarized. In the experiments, we used two types of small scale PWR hot legs. One was a 1/5th scale rectangular duct, and the other was a 1/15th scale circular pipe. Air and water were used for gas and liquid phases. The air flow rate and the supplied water flow rate were varied to observe flow pattern and measure the counter-current flow limitation (CCFL) characteristics. Flow patterns in the elbow and the inclined section were strongly affected by those in the horizontal section. In the 1/15th scale circular pipe experiments, CCFL characteristics obtained by increasing the air flow rate differed from those obtained by decreasing it. CCFL characteristics corresponded to the flow pattern transition. In the numerical simulations, we used a three-dimensional two-fluid model to evaluate the capability of predicting counter-current flow in the hot leg. Good agreements between measured and predicted flow patterns and CCFL characteristics were obtained by using an appropriate set of correlations for interfacial friction coefficient. We also carried out simulations of actual hot leg conditions to examine the effects of fluid properties and size. Predicted flow patterns and CCFL characteristics were close to those of scale model calculations. We concluded the combination of calculation model and interfacial friction coefficients used in this study can predict the counter-current flow in a hot leg.
Investigations on Friction Resistance of Forced Circulation in a Circular Tube Under Inclination and Rolling Conditions
Authors: Du Sijia , Zhang Hong
Published: April 2011

The friction resistance under marine conditions is very important to a marine reactor system in the ocean. The traditional empirical or semi-empirical relationships were used in thermo-hydraulic system programmes at the present time, while the applicability of these relationships under ocean conditions was not validated adequately. Experiments have been conducted to study the friction resistance of forced circulation when the test tube was at inclined attitude and rolling model. The experimental setup consists of stainless steel cylinder tube as test section with 10mm inside diameter and 1200mm length, is subjected to a constant wall heat flux boundary condition. The investigation covers Reynolds number range from 1.0e+4 to 1.2e+5, and the test section angles of inclination including 10°, 20° and 30°, amplitudes of rolling angle including 10°, 15° and 20° with the rolling periods of 10s, 15s and 20s. Pressure drop was measured to calculate the coefficient of friction resistance. CFD method was used to simulate the flow characteristics under inclination and rolling conditions, the results were compared with experimental data. The investigations showed that the influence of inclination and rolling to friction coefficient is not obvious.
Resonance Calculation for Large and Complicated Geometry Using Tone’s Method by Incorporating the Method of Characteristics
Authors: Hao Yu , Yamamoto Akio , Yamane Yoshihiro
Published: April 2011

Applicability and efficiency of Tone’s method used for resonance calculations is improved by incorporating the Method of Characteristics (MOC). Verification calculations are carried out in three configurations, i.e., infinite slab, infinite cylinder, and pin-cell. The validity of the modified Tone’s method has been confirmed through the results of verification calculations. Since MOC has excellent geometric flexibility and efficiency for large and complicated configurations, the proposed method can be applied to a large, complicated, and general geometry such as a whole fuel assembly or a full core configuration.
Irradiation Damage and Embrittlement in RPV Steels Under the Aspect of Long Term Operation: Overview of the FP7 Project LONGLIFE
Authors: Altstadt Eberhard , Bergner Frank , Hein Hieronymus
Published: April 2011

The increasing age of the European NPPs and envisaged lifetime extensions up to 80 years require an improved understanding of RPV irradiation embrittlement effects connected with long term operation (LTO). Phenomena which might become important at high neutron fluences (such as late blooming effects and flux effects) must be considered adequately in the safety assessments. Therefore the project LONGLIFE was initiated within the 7th Framework Programme of the European Commission. The project aims at: i) improved knowledge on LTO phenomena relevant for European reactors; ii) assessment of prediction tools, codes, standards and surveillance guidelines. In the paper, we give an overview of the project structure and the related tasks. Furthermore we present two examples for the experimental evidence of LTO relevant phenomena: the first example is related to the flux dependence of defect cluster formation in a neutron irradiated weld material. We have found that the size of the irradiation induced defects exhibits a flux effect whereas the mechanical properties are almost independent of the flux. The second example refers to the acceleration of irradiation hardening after exceeding a threshold fluence. This effect was observed by means of both small angle neutron scattering (SANS) and tensile testing for low Cu RPV steels irradiated at a temperature of 255 °C. These examples demonstrate that LTO irradiation effects have to be investigated in more detail to guarantee the applicability of the embrittlement surveillance guidelines beyond 40 years of operation.
Bubble Sliding Process on Subcooling Flow Boiling in Vertical Rectangular Narrow Channel
Authors: Yuan De-wen , Pan Liang-ming , Wei Jing-hua , Chen De-qi
Published: April 2011

In present work, a visualized investigation was conducted to understand the phenomenon of bubble sliding on vertical heating wall with upward subcooled flow boiling in narrow rectangle channel with gaps of 2.5 mm. Experimental results show that the characteristics of bubble slipping are significantly influenced by mainstream velocity and temperature. With increasing mainstream velocity, bubble sliding distance and speed are all increased. In present working conditions, the bubbles sliding distance is of 3–10 times of the bubble departure diameter. Based on the energy balance on bubble sliding process, the impact of bubble sliding on the characteristics of heat transfer was analyzed as well. Considering the different mechanisms effect on heat transfer, such as micro-layer evaporation, transient conduction and single-phase forced convection between fluid and wall, the ratio of every mechanism was calculated and compared. Analysis results show that the bubble with sliding movement have significant heat transfer enhancement than that of the stationary bubbles. With the proposed concept of disturbance intensity of sliding bubble, the heat flux expressions caused by disturbance of sliding bubble or stationary have been proposed.
The Welding Deformations of Stainless Steel Pipes With Thick Wall by Narrow Gap Gas Tungsten Arc Welding
Authors: Zhang Jainxun , Liu Chuan , Niu Jing
Published: April 2011

The austenitic stainless steel pipes with thick wall are widely used in the nuclear power station and are welded by narrow gap gas tungsten arc welding process. The welding deformations of multi-pass butt-welded pipes with 65 and 70mm thickness are investigated experimentally and numerically in the paper. The transient axial deformation and axis shift deformation are measured during welding. An axisymmetric FE model and a thermal mechanical calculating procedure are presented to simulate the welding axial deformations. An effective calculating method by only considering the contraction of each welding pass in the model is proposed. The experimental results show that the axis shift deformation is very small and demonstrates an elastic movement during welding; the axial shrinkage of welded pipes is the mainly deformation, which is very significant during the first several weld pass, and decreases sharply after the weld groove has been filled at the height of 30% wall thickness that makes the stiffness of pipes large enough to resist the welding shrinkage.
SuperCritical Water-Cooled Nuclear Reactor With Intermediate Heat Exchangers
Authors: Thind H. , Pioro I. , Harvel G.
Published: April 2011

At present, there are a number of Generation-IV nuclear reactor concepts under development worldwide, and the SuperCritical Water-cooled nuclear Reactor (SCWR) type is one of them. The main objective of developing SCWRs is to: 1) Increase the thermal efficiency of current Nuclear Power Plants (NPPs) from 30–35% to approximately 45–50%, and 2) Decrease capital and operational costs. SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25 MPa and outlet temperatures up to 625°C). This paper presents a SCWR single-reheat indirect cycle concept with intermediate heat exchangers. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, heat exchangers separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in the reactor building. The nuclear activities stay within the reactor building, and there is a reduced possibility for radioactive contamination of equipment in the turbine building. As SCW NPPs will have much higher operating thermal hydraulic parameters this paper analyzes the technical challenges and higher costs typically associated with heat exchangers. The double-pipe heat exchanger is analyzed in depth to determine the heat-transfer surface area, number of units and physical dimensions of the heat exchanger. This study will help to determine whether the advantages of the indirect cycle justify implementation of heat exchangers at a SCW NPP.
Advanced MOX and IMF Fuel Assembly Designs for Multi-Recycling Transuranics
Authors: Zhang Yunhuang , Ragusa Jean C.
Published: April 2011

Several new fuel assembly designs for multi-recycling Transuranics from spent nuclear fuel are proposed and investigated. Among these are (1) Mixed Oxide Fuel with Enriched Uranium (MOX-EU), in which Plutonium oxide and U-235 enriched Uranium oxide are mixed (2) MOX fuel with Americium coating, in which a thin layer of Americium is applied to the outer surface of the MOX fuel pellet, and (3) an heterogeneous fuel assembly consists of Inert-Matrix Fuel (IMF) pins at the periphery and UOX pins in the inner zone. All these designs are compatible with standard PWR utilizing 17×17 fuel assemblies. In-reactor fuel depletion simulation and long-term isotopic decay calculation are carried out using DRAGON[1] and ORIGEN[2], separately. Transuranics mass balance and long-term radiotoxicity analyses are implemented and the results are normalized to per 1TWh-electricity produced.
LBE-Water Interaction in LIFUS 5 Facility to Study a SGTR Event Under ELSY Reactor Conditions
Authors: Bernardi Davide , Ciampichetti Andrea , Tarantino Mariano , Coccoluto Giovanni , Forgione Nicola , Poli Francesco , Catanorchi Marco
Published: April 2011

In the framework of the research activities of the EURATOM FP6 project named ELSY (European Lead-cooled System), aimed at demonstrating the possibility of designing a competitive and safe fast critical reactor based on the Generation IV Lead Fast Reactor (LFR) concept, the study of the lead-water interaction following an incidental SGTR (Steam Generator Tube Rupture) event is an important issue to address. To simulate such event, an experimental test has been carried out on the LIFUS 5 facility at the ENEA Brasimone Research Centre, in order to assess the physical effects and the possible consequences connected to this kind of interaction. The experiment has been conducted by injecting water at the pressure of 185 bar and with a temperature of 300 °C into a volume of 80 l of Lead Bismuth Eutectic (LBE) kept at atmospheric pressure and at a temperature of 400 °C. The experimental facility has been suitably modified in order to reproduce as close as possible the operating conditions of the ELSY Steam Generator Unit (SGU), in which a free volume of cover gas (argon) is foreseen at the top of the system, with the objective to dampen the pressure waves inside the SGU itself. The experimental test has been supported through a numerical modelling campaign performed at the University of Pisa by means of the SIMMER code within both 2-D (SIMMER III) and 3-D (SIMMER IV) models. Pre-test simulations have been carried out to aid the design of the new facility configuration and to select the test conditions which could better reproduce the behaviour expected for ELSY. In addition, a post-test analysis has also been accomplished, allowing to compare the numerical and experimental results, so as to validate and assess the performance of the code when employed for this kind of applications.
A New Methodology to Obtain the 1D Cross-Sections for TRAC-BF1 Code: Application to Peach Bottom NPP
Authors: Barrachina T. , Miro´ R. , Verdu´ G. , Collazo I. , Gonza´lez P. , Concejal A. , Ortego P. , Melara J.
Published: April 2011

In TRAC-BF1 the cross-sections are specified in the input deck in a polynomial form. Therefore, it is necessary to obtain the coefficients of this polynomial expansion. One of the methods proposed in the literature is the KINPAR methodology developed in the UPV. This methodology uses the results from different perturbations of the original state to obtain the coefficients of the polynomial expansion. The simulations are performed using the SIMULATE3 code. In this work, a new methodology called SIMTAB-1D to obtain the cross-sections sets in 1D is presented. The first step consists of the application of the SIMTAB methodology, developed in the UPV, to obtain the 3D cross-sections sets from CASMO4/SIMULATE3. These 3D cross-sections sets are collapsed to 1D, using as a weighting factor, the 3D thermal and rapid neutron fluxes obtained from SIMULATE3. This new methodology will be applied to the simulation of the turbine trip transient in Peach Bottom NPP using the TRAC-BF1 code. The results of the steady state in TRAC-BF1 using the KINPA R methodology and the new methodology are compared with the reference SIMULATE3 results.
Introduction of Hot Isostatically Pressed, Reactor Coolant System Components in PWR Plant
Authors: Sulley John L. , Hookham Ian , Burdett Barry , Bridger Keith
Published: April 2011

This paper presents an overview of the work undertaken by Rolls-Royce to introduce Hot Isostatically Pressed (HIP) components into Pressurised Water Reactor plant, and also results of non-destructive and destructive examinations of a thick-walled pressure vessel. It presents the work from a design justification/manufacturing quality assurance perspective, rather than from a pure metallurgical perspective. Although the HIP process is not new, it was new in its application to Rolls-Royce designed nuclear reactor plant. As a consequence, Rolls-Royce has implemented an evolving, staged approach, starting with HIP bonding of solid valve seats into small bore valve pressure boundaries. This was followed by powder HIP consolidation of leak-limited, thin-walled toroids, and has culminated in the powder HIP consolidation of thick-walled components. The paper provides an overview of each of these stages and the approach taken with respect to justification. Mechanical testing and metallurgical examination results of sample material taken from different sections of a thick-walled component are presented. A full range of test results is provided covering, as examples: tensile, charpy and sensitization susceptibility. Differences in weldability between the HIPed and the previous forged form are also documented. The paper describes the benefits that Rolls-Royce has realised so far through the introduction of HIPed component. Structural integrity benefits are described, such as improved grain structure, mechanical properties, and ultrasonic inspection. Project-based benefits are also described, such as provision of an alternative strategic sourcing route, cost and lead-time reductions.
Application and Innovation of Equipment Reliability Management in Nuclear Power Stations
Authors: Zhang Ju F. , Chen Shi J. , Hong Jiang
Published: April 2011

Equipment reliability is a key factor to achieve excellent production capability and maintain high security for nuclear power stations. As the modern nuclear power industry develops in China, much effort is devoted to promoting the research on application of advanced equipment reliability management concepts and technologies. Besides the introduction and convergence of external advanced methods, some domestic innovations on equipment reliability technologies have also been accomplished recent years based on the long-term research and operating experience. These new ideas and technologies are fostered within the rapid development of China nuclear industry in the near decade, and aim at solving the emergent and common issues. As a typical representative of advanced equipment reliability management technology, INPO (Institute of Nuclear Power Operations) AP-913 equipment reliability process is accepted universally in global nuclear power industry and has been recommended by EPRI (Electric Power Research Institute), INPO and other authoritative organizations worldwide. It was originally introduced by many nuclear power stations in China as a reference model to establish their own equipment reliability management systems. As continuous research, attempts and ameliorations conducted, domestic innovations have been performed to develop more comprehensive and adaptive equipment reliability management technology, including integrating many existing reliability technologies, such as RCM, CCM, TCM, PFU, COMIS, and so on. This paper introduces some new research achievements on implementation of INPO AP-913 equipment reliability process at nuclear power stations in China. For each section contained in AP-913 process, the primary plans and suggestions proposed to meet the basic intent of AP-913 is firstly introduced to establish the practicable access for further development. And then, some improvement-step strategies and technologies are also presented to perform more sophisticated and effective management to improve equipment reliability in routine work. Finally, based on the actual conditions and demands of China nuclear power industry nowadays, some comprehension and advice beyond AP-913 process itself, which can also be considered as the localized modification, are presented. Moreover, the instance of AP-913 based equipment reliability management and development in Da Ya Bay Nuclear Power Station is also introduced as illustration. For Chinese nuclear power stations, these innovative attempts not only impelled the sustainable improvement of equipment management, but also exploited a feasible and compatible way to progress in Chinese nuclear industry pattern independently instead of seeking external support overseas.
Regulations Concerning Equipment Quality Supervision and Surveillance for Nuclear Power Plant EPC Projects
Authors: Zhang Liying
Published: April 2011

Use of the engineering, procurement and construction (EPC) contracting arrangement is increasing in China recently, especially for those important construction projects. The nuclear power plant construction projects are as well. Because of the associated hazards, nuclear power plant construction projects require more strict safety and quality requirements as compared with more conventional construction projects. As essential means and safety shields, the equipment quality supervision and surveillance requirements and their implementation mode should be explored to meet the requirements of EPC projects. It is highly demanded to establish new rules to meet the quick development of nuclear power plant construction projects. The article discusses the current equipment quality supervision and surveillance mode first, and then discusses the needs to build rules especially the significance of introducing independent organizations to supervise equipment quality for those nuclear power plant construction projects for using EPC contracting arrangement despite the supervision and surveillance conducted by manufactures, EPC contractors, nuclear power plant owners and the National Nuclear Safety Administration.
Preliminary Study of Large-Power PWR With Safety Balanced Design
Authors: Yang Jue , Zhu Xuenong , Fu Xiangang , Cai Wei , Ye Jie , Zhong Shan
Published: April 2011

Developing the advanced nuclear power plant design to meet the demanding safety, efficiency and environmental goals of electric utilities requires great efforts. In this paper, a design of the safety systems for the large-power PWR units is introduced, which is deemed a optimal combination of the passive safety systems with the active safety systems. The typical design basis accidents are analyzed for this safety system design, such as the Small Break LOCA, SGTR, SLB and Loss of Flow Accidents (LOFA). The results show that the safety systems of the passives combined the actives can mitigate effectively these typical accidents in large-power PWRs. PSA results also show that the passive safety systems contributes to the reduction of the CDF. It is preliminarily concluded that the passive combined active safety system is designed in balance.
RCCE-E 2005 Creator of Confidence Between Nuclear Safety Authority, Owner, and Suppliers
Authors: Haure J. M.
Published: April 2011

The paper deals with the design rules included in the RCCE code applicable to electrical, and instrumentation and control systems and components contributing to safety-class functions. It is a major document in the relationships between the Owner and its suppliers and the safety authorities. The code is periodically submitted to the Nuclear Safety Authority Assessment. Initially used for the French PWR program and published by the French Society for Design and Construction rules for Nuclear Island Components (AFCEN), the code evolves to comply with third generation PWR nuclear islands and aims to comply with national regulations if needed in addition to French regulations. It gathers within one set of rules, the design and construction practices of, the electrical and I&C systems and components, and installation engineering documents. Brief presentation of some items is proposed: - Relationships between Safety Analyses Report plant system and installation engineering; - Off-site and on-site sources requirements; - Periodic tests and permanent monitoring requirements to guarantee the safety function availability; - I&C architecture and Human Machine Interfaces; - Digital I&C systems requirements; - Establishing qualification to ambience conditions, that includes normal and mild conditions and, harsh and severe accident conditions. Custom-made approach based on the families of ambience defined as a combination of equipment mission time, and normal and abnormal conditions (pressure, temperature, radiation) enabling to cope better with environmental conditions; - High and Low frequencies disturbances protections (Lightning, GSM, DECT, WiFi); - Project data used by any supplier or designer such as site data, temperature of rooms, maximum design temperature of equipment, the voltages and frequency variation range and tolerances of the alternative current and direct current electrical networks, the decoupling data of environmental conditions; - Electrical equipment separation requirements and isolation and decoupling solutions designs; - Electrical enclosures design requirements. A conclusion on the evolution of AFCEN organization in charge of comprehensive set of technical codes on the design, construction and surveillance of civil work structures, fire protection, mechanical structures, the core and fuel design, the electrical and I&C equipment and systems of PWR nuclear island.
Estrogen Regulates Tumor Growth Through a Nonclassical Pathway that Includes the Transcription Factors ERβ and KLF5
Authors: Nakajima Yuka , Akaogi Kensuke , Suzuki Takashi , Osakabe Asami , Yamaguchi Chie , Sunahara Nanae , Ishida Junji , Kako Koichiro , Ogawa Sonoko , Fujimura Tetsuya , Homma Yukio , Fukamizu Akiyoshi , Murayama Akiko , Kimura Keiji , Inoue Satoshi , Yanagisawa Junn
Published: April 2011

Estrogens and antiestrogens exert their opposing effects on prostate tumor growth through ERβ and KLF5, but through distinct mechanisms.
Estrogen Receptor β: Switching to a New Partner and Escaping from Estrogen
Authors: Leung Yuet-Kin , Ho Shuk-Mei
Published: April 2011

ERβ promotes apoptosis in prostate cancer cells through an estrogen-independent mechanism.
Eosinophils and Metabolism
Authors: Mueller Kristen L.
Published: April 2011

Regulation of adipose tissue macrophages by eosinophils reveals an unexpected role for eosinophils in metabolic homeostasis.
Rac1 Regulates Both mTORC Complexes
Authors: Adler Elizabeth M.
Published: April 2011

Rac1 binds to mTOR, influencing the localization and activity of both mTORC1 and mTORC2.
A combined Fuzzy and Naïve Bayesian strategy can be used to assign event codes to injury narratives
Authors: Marucci-Wellman H , Lehto M , Corns H
Published: April 2011

Background

Bayesian methods show promise for classifying injury narratives from large administrative datasets into cause groups. This study examined a combined approach where two Bayesian models (Fuzzy and Naïve) were used to either classify a narrative or select it for manual review.

Methods

Injury narratives were extracted from claims filed with a worker's compensation insurance provider between January 2002 and December 2004. Narratives were separated into a training set (n=11,000) and prediction set (n=3,000). Expert coders assigned two-digit Bureau of Labor Statistics Occupational Injury and Illness Classification event codes to each narrative. Fuzzy and Naïve Bayesian models were developed using manually classified cases in the training set. Two semi-automatic machine coding strategies were evaluated. The first strategy assigned cases for manual review if the Fuzzy and Naïve models disagreed on the classification. The second strategy selected additional cases for manual review from the Agree dataset using prediction strength to reach a level of 50% computer coding and 50% manual coding.

Results

When agreement alone was used as the filtering strategy, the majority were coded by the computer (n=1,928, 64%) leaving 36% for manual review. The overall combined (human plus computer) sensitivity was 0.90 and positive predictive value (PPV) was >0.90 for 11 of 18 2-digit event categories. Implementing the 2nd strategy improved results with an overall sensitivity of 0.95 and PPV >0.90 for 17 of 18 categories.

Conclusions

A combined Naïve-Fuzzy Bayesian approach can classify some narratives with high accuracy and identify others most beneficial for manual review, reducing the burden on human coders.