bluuebunny/crossref_metadata_2025 huggingface dataset:
New Approach for Environmental Protection During Severe Accidents at Nuclear Power Plants
Authors: Kulyukhin S. A. , Mikheev N. B. , Falkovskii L. N. , Reshetov L. A. , Zvetkova M. Ya. , Osipov V. P. , Yagodkin I. V. , Skvortsov S. S. , Berkovich V. M. , Taranov G. S. , Grigor’ev M. M. , Noskov A. A. , Mitrofanov M. I. , Meshkov V. M.
Published: February 2010

This paper reports new approaches for increasing environmental protection during severe accidents at NPPs. For NPPs with two protective shells and pressure release system such as WWER-1000 we suggest a new comprehensive, passive-mode environmental protection system of decontamination of the radioactive air-steam mixture from the containment and the intercontainment area, which includes the “wet” stage (scrubbers, etc.), the “dry” stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For Russian WWER-440/V-230 NPPs we suggest three protection levels: 1) a jet-vortex condenser; 2) the spray system; 3) a sorption module. For modern designs of new generation NPPs, which do not provide for pressure release systems, we proposed a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable.
Contribution to the Design of Concrete Columns Reinforced With Externally Fiber Polymers
Authors: Prats Bella Ferran , Gonza´lez-Drigo Ramo´n , Bachiller San˜a Adrina
Published: February 2010

In the design of seismic category 1 buildings in nuclear power plants (NPP) or, outside the nuclear domain, in the conventional structural design of buildings, the seismic evaluation of these buildings may be done. In the occurrence of an earthquake in a NPP or in the case of changing the use of a conventional building, the seismic levels are modified. Then a new analysis need to be performed. This paper focusses on the situation where reinforcing the concrete building is needed and it also analyses how an extern reinforcement performed using polymers can be carried out to fulfill the new seismic requeriments. We present two main results: a) the resulting momentum-curvature diagrams obtained reinforcing standard segments embraced with polymers; b) the evaluation of the structure capacity on the basis of the modified diagrams. Finally, a modal pushover analysis is selected to perform the seismic evaluation of two types of concrete columns, those having a polymer reinforcement and those without it. This paper presents the basis of the subject in a theoretical form.
Electrical Network Modeling and Electrical Transfer Simulation of C.N. Asco´ I and C.N. Asco´ II, to Obtain Voltage and Frequency Limit Values Allowing the Electrical Transfer From the Main Generator to the External 110 kV Power Grid
Authors: Ban˜o´ Azco´n Alberto , Mollera Barriga Jose´
Published: February 2010

Broadly speaking, a simple electrical diagram of Asco´ I and Asco´ II power nuclear plants could be two power supply bus bars for general equipment classified No-1E, three power supply bus bars class No-1E for the Reactor Coolant Pumps (RCP) and two more bus bars classified 1E for safety related equipment. In normal operating mode, all the five power supply bus bars class No-1E are connected to the main generator (GP1) through two unit transformers (TAG1/2), while the two class 1E power supply bus bars are always connected to an external 110 kV power grid through two auxiliary transformer (TAA1/2). The main generator supplies power to an external 400 kV grid through the main transformer (TP1). The main circuit breaker is placed between the high voltage side of the main transformer and the 400 kV grid. With this configuration, the appearance of an abnormal condition that originates the trip of the main generator and the opening of the main circuit breaker from the external network, involves an electric transfer of the supply bus bars connected to the turbo generator to the external 110 kV power grid. The electric transfer to the external 110 kV power grid will be only possible if the frequency and voltage values are within the allowed range allowed by the grid’s protective relays. Two kinds of electrical transfers are possible: fast transfers and slow transfers. It will be necessary then to evaluate the limit values of voltage and frequency of the power grid that makes the electrical transfer possible in each case. In order to obtain the limit values previously mentioned, the electric system of the plant has been modeled. Different scenarios have been analyzed, taking into account the dynamic behavior of the system components and the delay of the protective relays actuation, verifying the electrical transfer for those situations. This analysis will give enough information to take the correct decisions for future design modifications, and it will assure that the electrical transfer will be done always with success.
Quantification of the RELAP5 Model Uncertainty for Application to the Loss-of-RHR Event During the Mid-Loop Operation
Authors: Kinoshita Ikuo , Nagumo Hiroichi , Yamada Minoru , Sasaki Yasuhiro , Yoshida Yoshitaka
Published: February 2010

Best estimate analysis method for the loss of Residual Heat Removal (loss-of-RHR) event during the mid-loop operation is being conducted along the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis method uses RELAP5/MOD3.2 as a best estimate analysis code. One of the important processes in the CSAU methodology is the development of the Phenomena Identification and Ranking Table (PIRT) which identifies thermal-hydraulic phenomena during the event and ranks the identified phenomena from the view point of influence on the safety evaluation parameters. The safety parameters for evaluation are Reactor Coolant System (RCS) pressure and reactor vessel water level. The PIRT for the reflux cooling of the loss-of-RHR event during the mid-loop operation was developed based on existing integral test results, plant analysis results and related papers considering influence on coolant distribution, non-condensible gas distribution and heat transfer. Referenced integral tests are ROSA-IV/LSTF, BETHSY, PKL and IIST. Uncertainty of RELAP5/MOD3.2 physical models related to high ranked phenomena identified in the PIRT for the reflux cooling is quantified using the related experimental data for application to PWR plant statistical analysis based on the developed verification matrix. Uncertainty quantified models are void model, horizontal stratified flow criteria and SG condensation heat transfer. These models are related to the following phenomena respectively. Void model (interfacial friction factor in bubbly and slug flow regimes): - Two phase expansion in core and upper plenum due to core boiling. - Two phase flow to Steam Generator (SG) inlet plenum and U-tubes. Horizontal stratified flow criterion: - Stratification of flow in hot leg. - Water transportation from hot leg to SG by steam flow. SG condensation heat transfer model: - Heat transfer in SG U-tube under presence of non-condensable gas. Distribution of model parameter multiplier which represents model uncertainty was obtained by either experiment analysis by RELAP5 or comparison of separate RELAP5 model prediction to experimental data. Mean value and standard deviation are calculated for distribution of model parameter multiplier.
Current Status and Performance of the J-PARC Accelerators
Authors: Sako Hiroyuki
Published: February 2010

J-PARC (Japan Proton Accelerator Research Complex) is a multi-purpose research facility for materials and life sciences, nuclear and particle physics, and nuclear engineering with extremely high power proton beams of 1 MW. The accelerator complex consists of a 400-MeV linac, a 3-GeV Rapid Cycling Synchrotron (RCS), and a 50-GeV Main Ring synchrotron (MR). Its goals are to provide MW-class beams at 3 GeV and at several 10 GeV, while it is a challenge to localize and suppress beam loss to the level to allow hands-on maintenance of accelerator components. The RCS scheme is adopted to realize them, which is advantageous over conventional Accumulation Ring (AR) regarding less beam loss problems due to lower beam current and easier construction and operation of a linac. RCS, however, required various challenging technologies such as ceramic ducts to reduce eddy current effects, high field Radio Frequency (RF) system, and paint injection technique (an injection scheme to reduce phase space density of the beam) to reduce space charge effects. The linac has also unique technologies to minimize beam loss, such as compact electromagnet Drift Tube Quadrupoles (DTQ’s) to control beam envelopes precisely, and a fast beam suspending system in Machine Protection System (MPS) with Radio Frequency Quadrupole linac (RFQ). The beam commissioning of the linac started in Nov. 2006, and its design energy of 181 MeV in the first construction phase was achieved in Jan. 2007. RCS beam commissioning started in Sep. 2007 and the beam was accelerated to the designed energy of 3 GeV in Oct. 2007. MR beam commissioning started in May 2008, and the beam acceleration to 30 GeV was established in Dec. 2008. The first neutron and muon beams were produced in May and Sep. 2008, respectively, at Materials and Life science experimental Facility (MLF). The linac commissioning has resulted in very stable beam with short down time. RCS commissioning quickly achieved beam acceleration and extraction, and paint injections are being studied intensively. RCS recorded the highest beam power of 0.21 MW in Sep. 2008 with beam loss well localized at the collimators. The linac beam energy will be upgraded to 400 MeV with Annular Coupled Structure linac (ACS) in order to increase the beam power to 1 MW. In the second construction phase, upgrade of the linac with 600-MeV Super-Conducting Linac (SCL) for Accelerator-Driven nuclear waste transmutation System (ADS) and upgrade of MR energy from 30 to 50 GeV are planned.
Sensitivity Analysis by the Use of a Surrogate Model During Large Break LOCA on ZION Nuclear Power Plant With CATHARE-2 V2.5 Code
Authors: Fouet Fabrice , Probst Pierre
Published: February 2010

In nuclear safety, the Best-Estimate (BE) codes may be used in demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters, which comes mainly from the lack of knowledge of the input parameters, is evaluated by estimating the 95% percentile with a high degree of confidence. IRSN, technical support of the French Safety Authority, develop a method of uncertainty propagation and chose to apply it to the calculation of the Peak Cladding Temperature (PCT) with CATHARE-2 V2.5 code during a Large Break (LB) LOCA event for ZION, a 4-loop PWR of Westinghouse design. As a general rule the Global Sensitivity Analysis (GSA) is done with linear correlation coefficients. This paper presents a new approach to perform a more accurate GSA to determine and to classify the main uncertain parameters: the SOBOL methodology. This technique requires simulating many sets of parameters to propagate uncertainties correctly, which makes of it a time-consuming approach. Therefore, it is natural to replace the complex computer code by an approximate mathematical model, called response surface or surrogate model. Kriging methodology (with simulated annealing optimization) for its construction and the SOBOL methodology for the GSA are used. The paper presents the application of the previously described methodology on a LB-LOCA scenario in ZION reactor, associated with 54 input parameters. The output is the first maximum peak cladding temperature of the fuel. Results show that the methodology could be applied to both high-dimensional complex problems and real nuclear power plant calculations.
Reliability Analysis by Multilevel Flow Models: A Qualitative Analysis Method and Its Application for FTA Validation
Authors: Yang Ming , Zhang Zhijian , Liu Jie , Yan Shengyuan
Published: February 2010

Fault Tree Analysis (FTA) is a powerful analytical technique for analyzing system reliability and safety by enumerating any possible safety-critical failure modes, which is very useful for identifying the risks and weaknesses in the system. Therefore, FTA is widely applied to the safety evaluation of large-scale and mission-critical systems. However, the following problems are usually pointed out when building a fault tree for a complex system: 1) System modeling is a hard and time consuming work, and 2) FTA models are difficult to be validated. In this paper, we propose a new method for system reliability analysis based on Multilevel Flow Models (MFM) and Goal Tree-Success Tree (GTST) methods. We use Goal Tree (GT) methodology to model the target system at a higher and system level, and use the Success Tree (ST) together with MFM at a lower and functional level. In this way, modeling effort could be significantly reduced. In this paper, an algorithm is also presented to translate the GTST-MFM model into ST model based on which qualitative reliability analysis can be performed by the Fusell-vesely algorithm. In this paper, a Low Head Safety Injection System (LHSIS) is taken as a case study to exemplify how to apply our proposed GTST-MFM method to model the system and to validate fault trees directly built by deductive method.
The Coupling of Nuclear Heat and Hydrogen Production Thermochemical Cycles
Authors: Wang Zhaolin , Naterer Greg , Gabriel Kamiel S. , Gravelsins Rob , Daggupati Venkata
Published: February 2010

The technology to use nuclear heat to thermally split water into hydrogen and oxygen attracts more and more attentions at present. This paper discusses some challenges to couple nuclear heat with thermochemical hydrogen production cycles. The challenges include matching the maximum heat grade of thermal chemical cycles and nuclear reactors, and extracting heat from nuclear reactors. Sulfur-iodine and copper-chlorine cycles are taken as typical examples for analysis and discussion. The heat grade and quantity required by each step of the cycles are discussed. The maximum heat grade of sulfur-iodine cycle is higher than 800°C which cannot be easily coupled by GenIV nuclear reactor and other sources of heat must be provided. In comparison, the maximum heat grade of copper-chlorine cycle is 530°C which can be coupled by more nuclear reactors such as advance Gen-III and future Gen-IV nuclear reactor. It is concluded that thermochemical cycles with lower temperature requirement are easier to couple with present and future generations of reactors. Low temperature thermochemical cycles such as copper-chlorine cycles are recommended to match the heat grade of most nuclear reactors. Some methods are proposed to couple heat between a thermochemical cycle and nuclear power generating station. Several heat extraction methods such as using working fluid of nuclear reactor to provide heat to thermochemical cycles are proposed in this paper.
Review of Recent Aging-Related Degradation Occurrences of Structures and Passive Components in U.S. Nuclear Power Plants
Authors: Nie Jinsuo , Braverman Joseph I. , Hofmayer Charles H. , Choun Young-Sun , Kim Min Kyu , Choi In-Kil
Published: February 2010

The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory (BNL) are collaborating to develop seismic capability evaluation technology for degraded structures and passive components (SPCs) under a multi-year research agreement. To better understand the status and characteristics of degradation of SPCs in nuclear power plants (NPPs), the first step in this multi-year research effort was to identify and evaluate degradation occurrences of SPCs in U.S. NPPs. This was performed by reviewing recent publicly available information sources to identify and evaluate the characteristics of degradation occurrences and then comparing the information to the observations in the past. Ten categories of SPCs that are applicable to Korean NPPs were identified, comprising of anchorage, concrete, containment, exchanger, filter, piping system, reactor pressure vessel, structural steel, tank, and vessel. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.
Evaluation of Load-Time Functions Due to Impact of Soft Missiles
Authors: Ruch Daniela , Herrmann Nico , Mu¨ller Harald S.
Published: February 2010

The threat spectrum of safety related structures has severely changed since September 11, 2001. The deterioration capability of large aircrafts impacting on structures has become aware in a dramatic manner. Simplified calculation models for aircraft impact are available but have not been verified for large commercial aircrafts. Based on this background, experimental studies are performed at the Materials and Testing Research Institute (MPA) Karlsruhe of the Universita¨t Karlsruhe (TH) to examine the shape and the peak loads of load-time functions induced by an impacting soft missile with a mass and stiffness distribution along the missile’s axis similar to a commercial aircraft onto a rigid target. Different levels of water filling for simulating the kerosene tanks can be realised to investigate their influence on the load-time function. The characterisation of the major impact phenomena and their variation due to changing test parameters is the foremost goal of these experiments. The missiles are accelerated up to the impact velocity using an air-pressure gun. Impact forces are measured with a piezoelectric measuring platform as well as with accelerometers positioned on the target. The impact is filmed using a high-speed video camera. The experimental setup, measurement results and their evaluation will be presented in detail. The experimental study forms the reference for accompanied numerical calculations using the explicit finite element code LS-DYNA. Thus, a numerical model validated on the experiments shall be generated in order to provide an enhanced confidence level for the future calculation of real impact scenarios. First simulation results will be shown and briefly compared to the solution using the Riera method. A possible enhancement of this analytical method based on the experiments is beyond the scope of the project.
Engineering Practices for Tokamak Window Assemblies
Authors: Waldon C. , Morrell R. , Buckthorpe D. , Davies M. , Sherlock P.
Published: February 2010

For fusion tokamak reactors the diagnostics and RF heating systems require the use of components with parts made of non-metallic materials. These can form part of the vacuum boundary of the tokamak which is the primary safety boundary and have a function of containing tritium fuel or activated gases and particulate debris. The engineering practices for such components and non-metallic materials are in an early state of preparation and require development to enable systems to be used in a safety and licensing context. Such developments will have to reflect the brittle nature of the materials, and are likely to be based on established arguments developed within the nuclear industry, such as containment and defence in depth. Given these requirements this task is a major challenge. The window systems fall broadly into two categories: • Transmission windows for the input of high-power microwaves to drive and heat the plasma; • Diagnostic windows to monitor the plasma. Currently there are no established fusion design codes that can be used to assure nuclear safety and a consistent engineering approach for either application. This paper reviews the progress made in developing such practices for transmission and diagnostic windows made from ceramic materials. The investigations undertaken and the engineering practices addressed for the tokamak windows generally fall into the following areas: • reviews of potential candidate materials along with a summary of the available property data; • definition of the function of torus window assemblies and an outline of the complexity and variety of design considerations (including historical failures, and statutory requirements); • development of the design methodology for technical ceramics; • definition of the design routes considered and selected (rule, analysis, experiment); • consideration of the material data available (or lack of) for technical ceramics and their failure criteria; • qualification and design of metallic / ceramic joints; • definition of the requirements with regard to quality control, from manufacture to in-service inspection; • development and formation of a draft code procedure. The practices and procedures developed are considered to be an important contribution and significant step forward in the development of a fusion tokamak windows code. Important contributions have been made to the design, procurement and installation philosophies for windows, especially the development of design criteria and the application of pressure proof-testing. This paper provides a review of key requirements and issues, with recommendations to allow development of the code for acceptance by nuclear regulators for tokamaks such as the International Tokamak Experimental Reactor (ITER) and future fusion reactor power plants.
Lessons Learned in Reviewing CCF Events
Authors: Mossa Ashley , Weston Rupert
Published: February 2010

The U.S. Nuclear Regulatory Commission (NRC) has an ongoing Common Cause Failure (CCF) data analysis program that periodically collects and evaluates information on component failures at U.S. commercial Nuclear Power Plants (NPPs). The primary information sources include the Licensee Event Reports (LER) and records from the Equipment Performance Information Exchange (EPIX) program. Once the information is collected, the failure records are evaluated to identify potential CCF events. CCF events are then coded, reviewed, and loaded into the NRC’s database. Verification of the CCF events is performed with the intended purpose of ensuring that events entered into the CCF database are indeed CCF events and that the event coding is consistent and correct. To ensure technical accuracy and correctness of the events loaded into the CCF database, the NRC requested the Pressurized Water Reactors Owners Group (PWROG) support in reviewing these events. Reviews of multiple data sets of CCF events were conducted on behalf of the PWROG. The data sets included CCF events that have occurred at U.S. commercial nuclear power plants. CCF events that occurred during 2006 through 2007 were included in the most recent data set that was reviewed. The level of information provided for reported CCF events varies from utility-to-utility. Without utility participation or input, the lack of consistency and varying level of detail can lead to incorrect interpretation and classification of a CCF event regarding its Probabilistic Risk Assessment (PRA) impact. This paper offers lessons learned from the reviews that were conducted. Insights for improving the consistency and level of detail related to the PRA information are summarized in this paper. The leading causes of initial misclassification of CCF events and patterns observed in conducting the reviews are discussed. The resolutions of misclassified CCF events are also discussed as part of the evaluation process to enhance the pedigree of the CCF database.
Advanced Application of BNCT in Advanced Cancers
Authors: Sztejnberg Manuel L. , Jevremovic Tatjana
Published: February 2010

We present a new concept of one form of radiation binary targeted therapy that may offer hope for the often fatal relapsed and/or metastasized HER2+ cancers. The idea is to deliver boronated (boron-10 isotope) anti-HER2 monoclonal antibodies (mAbs) to the patient to be deposited preferentially into the tumor followed by one session of a low energy neutron irradiation. Based on actual computed tomography data, we present the comprehensive theoretical (numerical) modeling of the new approach in designing the treatment conditions for the boron neutron capture therapy (BNCT) using the MITRII-FCB neutron beam facility. The results show the effectiveness of the proposed treatment option for the advanced breast cancers and the metastasized breast cancers in the lungs of a patient. Our theoretical analysis concludes that with a boron concentration of ∼316 μg/g in tumor and a tumor-to-healthy tissue boron concentration ratio of 35:1, this new BNCT breast cancer treatment can be performed with very low doses to normal tissue and 50 Gy, or higher, doses delivered to the tumor regions. In particular, when applied to the breast cancer treatment, the resulting doses to skin and lung remain under the tolerance dose values. We also went beyond the traditional application of the BNCT and analyzed its applicability in targeting the metastasized breast cancer; using the same theoretical approach we determined the doses delivered into the patient lung with scattered cancer loci.
Sensitivity Investigation of Specimen Shape on Elastic Behaviour of Perforated Plate
Authors: Ha Chang-Hoon , Park Tae-Jung , Kim Moo-Yong , Seon Kwang-Sang , Koo Jae-Mean , Seok Chang-Sung
Published: February 2010

There are various types of tube support plates installed in a steam generator according to the component designer’s preference. Most widely used types of tube support plates are BTSP (broached tube support plate), ATSG (advanced tube support grid), and the eggcrate. In this study, trefoil BTSP specimens made of ASME stainless steel are analyzed and tested. This study is to investigate the effect of specimen shape on an elastic behavior of trefoil BTSP through the compression and bending tests. Prior to the compression and bending tests of BTSP specimens, the equivalent elastic properties of BTSP unit cell are analyzed by the finite element analysis according to the different loading orientation as well as size of the model. Autodesk® Inventor™ software was used to make an analytical model and ANSYS® software was used for the finite element analysis and post-processing. Five and three different shapes of trefoil BTSP specimens are machined and utilized for the compression and bending (4-point and 3-point side bending) tests, respectively. Through the finite element analyses, compression, and bending tests, the equivalent elastic modulus of trefoil BTSP specimen is suggested to be 6,254MPa (907ksi) and the equivalent Poisson’s ratio as 0.64. Specifically the CS5 type specimen which has a ratio of one-fourth (= width/length) was revealed as an appropriate shape of specimen to show those elastic behavior.
Optimization of Dead-Layer Thickness for a HPGe Detector Using UCODE-MCNP Codes
Authors: Shaaban Noha , El Gammal Wael , Nasif Hesham
Published: February 2010

The use of modeling programs to predict the response of HPGe detectors is increasing in importance due to the extensive laboratory work, both in term of source preparations and measuring time. MCNP code is a powerful and useful tool for the simulation of Ge-detector efficiency calibration. The experimental efficiency data and MCNP calculations based only on the known physical measurements of the HPGe crystal do not agree well in some detectors. Detector construction materials and surface dead layers must be well specified. The dead layer of Ge detector is one of the most important factors that affect the calculations. In addition, and if provided by the manufacturer, the dead layer may changes with time. Consequently, it is necessary to optimize the thickness of the detector’s dead layer in order to obtain more accurate results for the efficiency of the detector using Monte Carlo calculations. Our approach consists of employing hybrid UCODE-MCNP codes to optimize the dead layer of the Ge-crystal aiming at decreasing discrepancies between experimental and simulated data of the Ge detector efficiency. UCODE has two attributes that are not jointly available in other inverse models: (1) the ability to work with any mathematically based model or pre- or post processor with ASCII or text only input and output files, and (2) the inclusion of more informative statistics.
Experimental Investigation of Non-Condensable Gases Effect on Novovoronezh NPP-2 Steam Generator Condensation Power Under the Condition of Passive Safety Systems Operation
Authors: Kopytov I. I. , Kalyakin S. G. , Berkovich V. M. , Morozov A. V. , Remizov O. V.
Published: February 2010

The design substantiation of the heat removal efficiency from Novovoronezh NPP-2 (NPP-2006 project with VVER-1200 reactor) reactor core in the event of primary circuit leaks and operation of passive safety systems only (among these are the systems of hydroaccumulators of the 1st and 2nd stages and passive heat removal system) has been performed based on computational simulation of the related processes in the reactor and containment. The computational simulation has been performed with regard to the detrimental effect of non-condensable gases on steam generator (SG) condensation power. Nitrogen arriving at the circuit with the actuation of hydroaccumulators of the 1st stage and products of water radiolysis are the main sources of non-condensable gases in the primary circuit. The feature of Novovoronezh NPP-2 passive safety systems operation is that during the course of emptying of the 2nd stage hydroaccumulators system (HA-2) the gas-steam mixture spontaneously flows out from SG cold headers into the volume of HA-2 tanks. The flow rate of gas-steam mixture during the operation of HA-2 system is equal to the volumetric water discharge from hydroaccumulators. The calculations carried out by different integral thermal hydraulic codes revealed that this volume flow rate of gas-steam mixture from SG to the HA-2 system would suffice to eliminate the “poisoning” of SG piping and to maintain necessary condensation power. In support of the calculation results, the experiments were carried out at the HA2M-SG test facility constructed at IPPE. The test facility incorporates a VVER steam generator model of volumetric-power scale of 1:46. Steam to the HA2M-SG test facility is supplied fed from the IPPE heat power plant. Gas addition to steam coming to the SG model is added from high pressure gas cylinders. Nitrogen and helium are used in the experiments for simulating hydrogen. The report presents the basic results of experimental investigations aimed at the evaluation of SG condensation power under the inflow of gas-steam mix with different gases concentration to the tube bundle, both under the simulation of gas-steam mixture outflow from SG cold header to the HA-2 system and without outflow. As a result of the research performed at the HA2M-SG test facility, it has been substantiated experimentally that in the event of an emergency leak steam generators have condensation power sufficient for effective heat removal from the reactor provided by PHR system.
Point Lepreau Refurbishment Project Level 2 PSA Applications
Authors: Comanescu L. , Sawh S. , Wei M. , Lee B. , Petrescu A. , Nainer A. , Jaitly R. K. , Jean A. F. , Basque D. F. , Mullin D. S.
Published: February 2010

A Probabilistic Safety Assessment (PSA) for Point Lepreau Generating Station has been completed as part of the plant Refurbishment Project. The main objective of this PSA is to provide insights into plant safety design and performance, including the identification of dominant risk contributors and assessing options for reducing risk. The scope of this assessment covers Level 1 and 2 PSA and includes internal events for full power and shutdown, internal fires and internal floods as well as PSA-based seismic margin assessment (SMA) for full power operation. Following the accident sequence quantification for internal events, fire and flood, the results were integrated to provide an overall estimation of the Severe Core Damage Frequency (SCDF) and the Large Release Frequency (LRF) for the refurbished Point Lepreau plant. Importance analysis was performed on the integrated results to identify risk-significant failures, using Fussell-Vesely and Risk Achievement Worth indices, and risk-contributors using Risk Reduction Worth indices. Based on the importance measures, analysis was performed to evaluate the sensitivity of the SCDF and LRF results to the dominant contributors. Uncertainty analysis was also performed to provide qualitative discussions and quantitative measures of the uncertainties in the results of the PSA, namely the frequency of severe core damage or external releases. Based on the results, recommendations were made to improve maintenance, testing, training procedures as well as housekeeping. Also, results of the Level 1 and 2 PSA have been used to determine the safety important systems and components. This paper discusses the key results and recommendations of the Level 2 PSA as well as the methodology used to determine the safety important systems.
Westinghouse AP1000 Solution to Long-Term Cooling Debris Concerns
Authors: Schulz Terry L. , Andreychek Timothy S. , Song Yong J. , McNamee Kevin F.
Published: February 2010

The AP1000 is a pressurized water reactor with passive safety features and extensive plant simplifications that provides significant and measurable improvements in safety, construction, reliability, operation, maintenance and costs. The design of the AP1000 incorporates a standard approach, which results in a plant design that can be constructed in multiple geographical regions with varying regulatory standards and expectations. The AP1000 uses proven technology, which builds on more than 2,500 reactor years of highly successful Westinghouse PWR operation. The AP1000 received Final Design Approval by the Nuclear Regulatory Commission in September 2004. The AP1000 Nuclear Power Plant uses natural recirculation of coolant to cool the core following a postulated Loss Of Coolant Accident (LOCA). Recirculation screens are provided in strategic areas of the plant to remove debris that might migrate with the water in containment and adversely affect core cooling. The approach used to avoid the potential for debris to plug the AP1000 recirculation screens is consistent with the guidance identified in Regulatory Guide 1.82 Revision 3, the Pressurized Water Reactor (PWR) Industry Guidance of NEI 04–07, and the Nuclear Regulatory Commission’s Safety Evaluation on NEI 04–07. Various contributors to screen plugging were considered, including debris that could be produced by a LOCA, resident containment debris and post accident chemical products that might be generated in the coolant pool that forms on the containment floor post-accident. The solution developed for AP1000 includes three major aspects, including the elimination of debris sources by design, features that prevent transportation of debris to the screens and the use of large advanced screen designs. Measures were taken to design out debris sources including fibers, particles and chemicals. Available industry data from walkdowns in existing plants is used to determine the characteristics and amounts of the fibrous and particulate debris that could exist in containment prior to the LOCA. Materials used in the AP1000 containment are selected to eliminate post accident chemical debris generation. Large, advanced screen designs that can tolerate significant quantities of debris have been incorporated. Testing has been performed which demonstrates that the AP1000 screens will have essentially no head loss considering the debris that could be transported to them. Testing has also been performed on an AP1000 fuel assembly that demonstrates that it will also have essentially no head loss considering the debris that could be transported to it.
Point Kinetics LOCA Scenarios Assessment at the Using of Hafnium Absorber in Advanced CANDU Fuel Designs
Authors: Prodea Iosif , Ma˘rgeanu Cristina Alice , Prisecaru Ilie , Da˘nila˘ Nicolae
Published: February 2010

The goal of the paper is to compare the results obtained through point kinetics calculations at the simulation of several CANDU® LOCA scenarios in the case of using Hafnium as burnable absorber in advanced CANDU fuel designs (ACR™-1000 based) along with standard CANDU NU fuel. The 3D DIREN_MG diffusion code developed in INR Pitesti was used to accomplish this task after upgrading by original procedures, which allow for Neutronics and ThermalHydraulics coupled calculations. The intervention of SDS1 was modeled until the transient is terminate by SDS1 action. The CANDU community has always interested in the void reactivity reducing-the major drawback of this power reactor. The paper’s novelty elements consist in the using of a newer WIMS code version (WIMSDB5 from NEA Data Bank) and an updated IAEA WIMS library based on newly ENDF/B-VII version along with a simplified DIREN_MG ACR-1000 core model. The results emphasized the critical Hafnium shell thicknesses which allow for a slightly negative cell Void Reactivity (VR) target and the direct consequences in the full core safety analysis.
ACR-1000™ Level 1 PSA Summary
Authors: Menon U. , Aprodu S. , Bettig R. , Henderson R. , Ha T. , Iliescu P. , Karimi B. , El-Mansi M. , Robinson J. , Torabi K. , Eman E. , Wu J. , Santamaura P.
Published: February 2010

The ACR-1000™ developed by Atomic Energy of Canada Limited (AECL) is a 1200 MWe - pressure tube type, light-water-cooled and heavy-water-moderated reactor, which has evolved from the well-established CANDU™ line of reactors. It retains the basic proven CANDU design features while incorporating innovations and state-of-the-art technologies to ensure fully competitive safety, operation, performance and economics. The major innovation in the ACR-1000 is the use of slightly enriched uranium fuel and light water coolant. ACR-1000 is a four-quadrant design (for easier maintenance and improved reliability). There are five safety systems in the ACR-1000; (i) two independent, diverse and fast acting shutdown systems (SDS1 and SDS2), which are physically and functionally independent from each other and from the reactor regulating system; (ii) Emergency core cooling system; (iii) Emergency Feedwater system; and (iv) Containment system, which includes a strong steel-lined containment structure. In addition the Reserve water system provides feedwater to the heat transport system, steam generators, moderator and shield cooling system for beyond design basis accidents. The Level 1 Probabilistic Safety Assessment (PSA) is conducted in support of the design phase of the ACR-1000. The purpose of Level 1 PSA is to identify whether the ACR-1000 design targets and the regulatory safety goal for severe core damage frequency (SCDF) are met with adequate margin and provide design feedback. An interim Level 1 PSA was conducted for internal at-power events. Interim assessments were conducted for shutdown state, internal fire and flood at-power events. An interim seismic margin assessment was conducted for the seismic events. The Level 1 PSA results show that the ACR™ design targets and safety goal for SCDF are met with significant safety margin. Based on the ACR-1000 Level 1 PSA, the accident behaviours of the ACR-1000 are well understood and their consequences can be predicted with a high-level of confidence. It also provides sufficient assurance that the release based regulatory safety goals are achievable for ACR-1000. The Level 1 PSA results also signify a robust design that provides a strong foundation for the ACR-1000 design. The paper summarizes the Level 1 PSA program, methodology followed, the results obtained, and insights gained during the development of the ACR-1000 Level 1 PSA.
Critical Velocity of Potential Flow in Interaction With a System of Plates
Authors: Kerboua Y. , Lakis A. A. , Thomas M. , Marcouiller L. , Toorani M. H.
Published: February 2010

The elastic structures subjected to flowing flow can undergo the excessive vibrations and consequently a considerable change in their dynamic behavior, and they may lose their stability. A fluid-solid finite element model is developed to model a set of plates subjected to flowing fluid under various boundary conditions, fluid level, and fluid velocity that strongly influence the dynamic behavior of the plates. A hybrid method, which combines the finite element approach with the classical theory of plates, is used to derive the dynamic equations of the coupled fluid-structure system. The membrane and the transversal displacement fields are modeled, respectively, using the bilinear polynomials and the exponential function. The structural mass and rigidity matrices are derived by exact integration of developed displacement field. The fluid pressure is expressed by inertial, Coriolis and centrifugal fluid forces written, respectively, as a function of acceleration, velocity and transversal displacement. The fluid dynamic pressure is determined using the potential flow equation. Integrating this dynamic pressure in conjunction with the structural element results in the flow-induced mass, damping, and stiffness matrices, hence, one can establish the dynamic equations of coupled fluid-structure system. The impermeability condition that ensures the permanent contact between the shell and the fluid is applied at the contact surface. A parametric study has been performed to investigate the effect of physical and geometrical parameters (e.g. boundary conditions, fluid level, and flow velocity) on the dynamic response of the coupled system. The results are in satisfactorily agreement with those of experiments and other theories.
Intelligent Infrastructure for Energy Efficiency
Authors: Gershenfeld Neil , Samouhos Stephen , Nordman Bruce
Published: February 2010

A substantial fraction of wasted energy can be recovered by extending insights from the architecture of the Internet to the infrastructure of buildings.
Noise Can Induce Bimodality in Positive Transcriptional Feedback Loops Without Bistability
Authors: To Tsz-Leung , Maheshri Narendra
Published: February 2010

Bursty, Infrequent Noise

In gene regulatory networks, positive feedback loops can give rise to bistability and hysteresis in gene expression, thereby allowing switching mechanisms and memory effects.
To and Maheshri
(p.
1142
; see the Perspective by

Levens and Gupta

) eschew the commonly held idea that sigmoidal promoter responses are required to achieve a steady-state bimodal response in a positive feedback loop. Instead, using a model and data from an experiment, they favor noisy gene expression and multiple, noncooperative transcription factor binding as an explanation for the bimodal response, and they expect that similar noisy systems are widespread in biology.
Individuals and the Variation Needed for High Species Diversity in Forest Trees
Authors: Clark James S.
Published: February 2010

Difference, Not Diversity

In tropical forests, as in the ocean plankton, thousands of species may compete for the same resources. How they succeed in coexisting remains one of the central paradoxes in the study of biodiversity. Theory shows that coexisting species must partition the environment, but such partitioning is not obvious. Using data from coexisting forest trees in the southeastern United States,
Clark
(p.
1129
) show that individual variation between members of the same species allows them to avoid direct competition: One plant may differ significantly from another in its requirements for light, nutrients, or moisture, yet remain within the general spectrum of features displayed by its conspecifics.
Plant Peptides Govern Terminal Differentiation of Bacteria in Symbiosis
Authors: Van de Velde Willem , Zehirov Grigor , Szatmari Agnes , Debreczeny Monika , Ishihara Hironobu , Kevei Zoltan , Farkas Attila , Mikulass Kata , Nagy Andrea , Tiricz Hilda , Satiat-Jeunemaître Beatrice , Alunni Benoit , Bourge Mickael , Kucho Ken-ichi , Abe Mikiko , Kereszt Attila , Maroti Gergely , Uchiumi Toshiki , Kondorosi Eva , Mergaert Peter
Published: February 2010

Legume Symbiosome

Leguminous plants (peas and beans) are major players in global nitrogen cycling by virtue of their symbioses with nitrogen-fixing bacteria that are harbored in specialized structures, called nodules, on the plant's roots.

Van de Velde
et al.

(p.
1122
) show that the host plant,
Medicago truncatula
produces nodule-specific cysteine-rich peptides, resembling natural plant defense peptides. The peptides enter the bacterial cells and promote its development into the mature symbiont. In a complementary study,

D. Wang
et al.

(p.
1126
), have identified the signal peptidase, also encoded by the plant, that is required for processing these specialized peptides into their active form.
Observing Weather from Space
Authors: Kidder Stanley Q. , Vonder Haar Thomas H.
Published: February 2010

Since the first weather satellite was launched 50 years ago, satellite observations have revolutionized weather forecasting.
Of HeLa and Human Lives

The Immortal Life of Henrietta Lacks by
Rebecca Skloot

Crown, New York, 2010. 392 pp. $26, C$32. ISBN 9781400052172.
Authors: Boerner Leigh Krietsch
Published: February 2010

Ethical and racial issues in medicine and biomedical research are among the facets of Skloot's spellbinding story of Henrietta Lacks, the HeLa lineage cultured from her cancer, and how that legacy affected her family.
Seawater Chemistry and Climate
Authors: Elderfield Harry
Published: February 2010

Reconstructions of past seawater chemistry provide insights into the driving forces behind long-term climate change.
Changing Views of the San Andreas Fault
Authors: Scharer Katherine
Published: February 2010

A combination of high-resolution laser imaging with improved radiocarbon dating techniques is providing new ways to view earthquake behavior.
Reliable Noise
Authors: Levens David , Gupta Ashutosh
Published: February 2010

Assessing how the noise created in transcription factor regulatory circuits affects gene expression is essential to understanding network operation and output.
Turning Off Inflammation Signaling
Authors: Sriskantharajah Srividya , Ley Steven C.
Published: February 2010

A cytosolic protein controls interactions between key ubiquitin-conjugating enzymes, thereby regulating the expression of proinflammatory genes.
The Prediction Puzzle
Predicting the Unpredictable
The Tumultuous Science of Earthquake Prediction

by Susan Hough

Princeton University Press, Princeton, NJ, 2010. 269 pp. $24.95, £16.95. ISBN 9780691138169.
Authors: Scholz Christopher H.
Published: February 2010

Hough's accessible history of earthquake prediction explores why seismologists remain unable to warn when major quakes will strike.
New Network to Track Drugs and Vaccines in Pregnancy
Authors: Couzin-Frankel Jennifer
Published: February 2010

A new effort to nail down the risks of either using or doing without key medications during pregnancy is being launched this week with $12.5 million from two U.S. agencies.
Century-Long Debate Over Momentum of Light Resolved?
Authors: Cho Adrian
Published: February 2010

Physicists have been debating the formula for the momentum of light zipping through a transparent material ever since two different formulas were proposed more than 100 years ago. Now a theorist says both formulas are correct, but they denote different things and apply in different contexts.
Experts Map the Terrain of Mood Disorders
Authors: Holden Constance
Published: February 2010

Authors of the latest revision of psychiatry's
Diagnostic and Statistical Manual of Mental Disorders
(
DSM-V
) are debating whether the relationship between anxiety and depression is so close that they should be subsumed into a supercategory of human hopelessness, fear, and existential angst.
Scientists Grapple With 'Completely Out of Hand' Attacks on Climate Science
Authors: Kintisch Eli
Published: February 2010

A symposium organized at the last minute at the annual meeting of the American Association for the Advancement of Science by two of the world's most prominent scientific organizations addressed recent attacks on an increasingly beleaguered climate science community.
The Latest on Geoengineering
Authors: Kintisch Eli
Published: February 2010

Preliminary findings presented at the annual meeting of the American Association for the Advancement of Science suggest that some proposed techniques to cool the planet manually may have fewer barriers than previously thought.