bluuebunny/crossref_metadata_2025 huggingface dataset:
Thermohydraulic Studies of a New Passive Concept for Interim Storage: Experiment and Analysis
Authors: Duret Bernard , Argoud Jean Claude , Sublon Judicael
Published: February 2010

Concerning the radioactive waste management, thermoaeraulic of a new horizontal and passive concept in a closed space is studied. Corrosion is avoided, as heat transfer use confined air in natural convection. An experimental real scale mock-up has been carried out to demonstrate feasibility of this new concept and to provide reliable data for the validation of numerical simulation methods. The design of such facilities implies thermal-aeraulic studies in order to predict air and wall temperatures. For that reason, numerical CFD tools have to be validate in order to model the cooling of spent nuclear fuel in a real interim storage and make sure that safety thresholds never exceed critical values. The experimental loop represents a slice of the real module at the same scale in height and length. It consists of four superposed horizontally annular spaces (length: 5m; air gap: 5cm) where inner cylinder is equipped with electrical wires, aiming to reach a uniform and controlled power density. A natural air circulation appears between the tubes and a special cooled roof. The whole system is inside an air-conditioned building. For each annular space, inner surfaces are painted in black with a known emissivity. Instrumentation includes thermocouples in the air flow channel on the cylinders and on the roof wall, and air velocity measurement (hot wire transducer and Laser Doppler Velocimetry). During experiments, the external temperature and the heating power have been controlled and adjusted to simulate various thermal conditions. Numerous configurations and geometrical parameters have been studied to optimize the cooling conditions. The main experimental results obtained with different values of the parameters like the kind of cooling (water or air), power density (500 to 750W/m2) are presented as well as the heterogeneous power distribution on the four channels. Steady-state local temperatures are presented, with air flow distribution within annular space. Using this database, a special CFD tool has been developed to model complex and coupled phenomena. Then, numerical simulations results have been compared on a reference test case, firstly with water cooling and secondly with natural air cooling. Experimental and numerical data are presented and analysed. From that numerical validation, thermohydraulic model’s extrapolation is now in progress to design a real interim storage with 49 heated sub-channels and propose power loading in order to respect safety rules, especially temperature criteria on the wall.
Analysis of IHX Vertical Position Effects in the Reactor Vessel of KALIMER-600
Authors: Han Ji-Woong , Eoh Jae-Hyuk , Lee Tae-Ho , Kim Seong-O
Published: February 2010

In this research the vertical position effects of an IHX (Intermediate Heat eXchanger) during a loss of normal heat sink accident were investigated by varying the distance between the thermal centers of core and IHX. The conceptual design data of KALIMER(Korea Advanced Liquid Metal Reactor)-600(Hahn et al., 2007) was utilized as the reference reactor design. The thermal-hydraulic calculations for the steady and transient state have been done using the COMMIX-1AR/P code(Garner et al., 1992). In order to evaluate the effects of the IHX vertical position, the elevation of the IHX was increased by 1m and 2m respectively in comparison to the reference design and the required numerical analysis was completed. Several coastdown flow conditions were also considered to investigate the variation of the IHX elevation effect. The thermo-hydrodynamic behaviors in a steady and transient state were compared during early-phase of an accident. It was found that the IHX elevation enhances the circulation flow rate and contributes to the decrease of early-phase peak temperature for non-reverse flow coastdown conditions. IHX elevation also contributed to the extension of the non-reverse flow coastdown flow time range.
Development of Salt Particle Collection Device to Prevent SCC of Canisters
Authors: Takeda Hirofumi , Saegusa Toshiari
Published: February 2010

A natural cooling system is economical for removing the decay heat from casks at interim storage facilities of spent nuclear fuel. At storage facilities of concrete casks built near the seashore, the air including the sea salt particles comes into the concrete casks and may cause Stress Corrosion Cracking (SCC) to the canister made of welded stainless steel plates. In order to prevent SCC on the canister, it is necessary to keep the density of salt on the surface of the canister smaller than the threshold which causes SCC. In this study, the authors propose a salt particle collection device with a low flow resistance which doesn’t block the air inlet of the storage building. The salt particle collection device is installed at the inlet and composed of a duct with multiple trays where the water is filled. When the air including sea salt particles comes through the duct, it collides with the surface of the water in the trays, and the part of sea salt particles in the air dissolves in the water. Therefore, the salt carried into the building by the air is reduced. The salt water in the trays is discharged out of the building by overflowing. The device has the following characteristics. a) Because of the low flow resistance, the device doesn’t block the inflow of the air which needs for removing the heat from casks. b) Rainwater may be usable for the water used in the device. c) Because of the simple structure, the maintenance is easy. The authors conducted the experiments using the device in the laboratory and outdoors near the seashore. The obtained results are as follows: (1)The pressure loss of the device is 1/7 of that of a filter used in a forced cooling system and the efficiency of salt particle collection is approximately 24%. (2)The efficiency of salt particle collection in the outdoor tests is larger than that in the laboratory tests. By further experimental study, the authors will develop the device with lower pressure loss and higher efficiency of salt particle collection.
ABWR Start-Up Test Analysis With Transient Code BISON
Authors: Yano Kazuki , Hayashi Yamato
Published: February 2010

New ABWR plants are planned or studied in Japan as well as the U.S. and other countries. Multiple safety analyses have been performed for ABWR in order to demonstrate the safe and stable performance of this type of plants. Westinghouse code package (BISON, PHOENIX4 and POLCA7) has been used for most BWR plants in Europe and the U.S., while it has not been used for ABWR transient analyses yet. In order to verify the applicability of BISON and POLCA7 code to ABWR, input data for these codes have been generated in accordance with geometry and detailed design specifications, such as control system response and set points of various transient mitigation functions. The calculated steady-state and transient results are compared to the recorded test data of past ABWR start-up tests performed prior to commercial operation. For the validation of BISON code, control system tests (pressure, recirculation flow, feedwater flow), feedwater pump trip, reactor internal pump trip, load rejection and MSIV closure tests have been simulated and the obtained results show an overall good agreement with actual plant behavior. This validation analysis used output from POLCA7 as the input of kinetics with verified calculation for core performance of steady-state. Westinghouse code package was validated for applying to ABWR by compared with the data collected from startup test. It is expected to be capable for simulating transients of the reactor, therefore, it is adequate enough to design and predict of transient behavior for ABWR.
Analysis of ABWR Critical Control Operation by TRACG Code
Authors: Ishii Yoshihiko , Kitou Kazuaki , Ikegawa Tomohiko , Hasegawa Shin , Ochi Hitoshi
Published: February 2010

Hitachi utilized three-dimensional transient analysis to design and verify a critical-control mode algorithm of an automatic power regulator (APR). TRACG has a three-dimensional neutron kinetics model based on diffusion theory and a six-equation two-phase flow model. To verify the APR critical-control mode algorithm, an external-neutron-source model that makes possible to simulate a sub-critical initial core, and an APR system model were developed and added on TRACG. The code was verified by comparison of measurements and calculation results of ABWR start-up operation under the critical-control mode. The modified TRACG could simulate neutron count rates of start-up-range neutron monitors (SRNMs), reactor period, control rod operation timing, CR withdrawal length, and time of criticality declaration, well.
Variation of Critical Heat Flux by Flow Oscillation in a Small Vertical Channel
Authors: Murakami Toshihiro , Takei Rei , Okawa Tomio
Published: February 2010

The effect of sinusoidal oscillation of inlet mass flux on the critical heat flux (CHF) in forced convective boiling was investigated in experiment and numerical calculation. In the experiment, the test section was a small stainless steel round tube of 5 mm in inside diameter, filtrated and deionized tap water was used as a test fluid, and the flow direction was set to vertical upward. The heated length was 1,600 mm. Electric power supplied to a circulation pump was varied periodically to oscillate the inlet mass flux sinusoidally. Direct current was passed through the test section tube to heat it ohmically. The occurrence of critical heat flux condition was detected using the signal from the thermocouples that were spot-welded on the outer wall of the test section tube. In the present experimental conditions, it was expected that the critical heat flux condition was triggered by the dryout of liquid film in annular two-phase flow regime. The main experimental parameters were the time-averaged inlet mass flux and the amplitude and period of flow oscillation. The system pressure was also used as an important experimental parameter since a boiling water reactor is operated under high pressure condition. If the oscillation period is long enough, it is expected that the critical heat flux under the flow oscillation condition is close to that for the steady state when the flow rate is equal to the minimum flow rate in the oscillatory condition. On the other hand, the decrease of the critical heat flux would be mitigated if the oscillation period is shortened, since interaction would take place between the thin and thick film regions within a boiling channel. In accordance with this expectation, the critical heat flux measured under the flow oscillation condition was reduced with an increase in the oscillation period. It was demonstrated that the reduction of critical heat flux under flow oscillation condition can be correlated fairly well using the concept of dimensionless heated length. Numerical calculations using a one-dimensional three-fluid model were also carried. The calculated critical heat fluxes for flow oscillation conditions increased with increased value of dimensionless heated length, as in the present experiment.
Theoretical Calculation of the Critical Heat Flux in Annular Upward Flow in Vertical Narrow Rectangular Channels
Authors: Du Dongxiao , Su Guanghui , Qiu Suizheng
Published: February 2010

The present paper has developed a mathematical separated flow model for annular upward flow in vertical narrow rectangular channels to predict the critical heat flux. The theoretical model is based on fundamental conservation principles: the mass, momentum, and energy conservation equation of liquid film and the momentum conservation equation of vapor core together with a set of closure relationships (such as entrainment rate, deposition rate, interfacial shear stress and initial entrainment fraction at the onset of annular flow). The predicted results are compared with the experimental data and fairly good agreement between them is achieved. By numerically solving the equations, liquid film thickness, liquid film velocity in the liquid film and heat transfer coefficient are obtained. With the applications of the present model, the critical heat flux in the rectangular channel is calculated and analyzed.
Differences in the Heat Transfer at the Core-Reflector Boundary of the PBMR Due to the Use of Different Numerical Methods
Authors: du Toit Pieter S. , Ubbink Onno
Published: February 2010

The PBMR (Pebble Bed Modular Reactor) is a High-Temperature Gas-cooled Reactor (HTGR) concept. One of the exercises of the PBMR benchmark of the Organization for Economic Cooperation and Development (OECD) is a steady state two-dimensional (2D) thermal-hydraulics simulation of a simplified PBMR with prescribed heat sources. Two different programs were used to model this exercise. They predicted similar core temperatures but the side reflector temperatures next to the core differed by more than 30 °C (when using a relatively coarse mesh). The underlying methods define temperatures at either vertices (VC) or at mesh cell centres (CC). A study was undertaken using one-dimensional (1D) implementations of the VC and CC methods to model a horizontal slice through the core. This study revealed the root cause of the different predictions. A modified version of the 1D CC method was developed that essentially predicts the same temperatures as the VC method. The extension of the modified method to two dimensions is under investigation. If the difference in predicted temperatures next to the core can be eliminated or reduced, then the focus can shift to other differences between the results of the two programs.
Experimental ECCS Sump Strainer Head Loss Testing and the Incorporation of CFD Computed Source Terms for Pressurized Water Reactors
Authors: Cain Stuart , Gartland Fariba , Bleigh James , Johansson Andrew , Schowalter David
Published: February 2010

High Energy Line Breaks (HELBs) inside nuclear reactor containment are recognized as challenges to Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear power plants arising from the collateral damage due to insulation, fireproofing, coatings, and other miscellaneous materials which are shredded and transported during the event. These materials, as well as latent debris (dirt and dust) will be transported towards the containment floor and the recirculation sump screens by flow from both the HELB and the containment spray headers. This debris, if washed towards the recirculation pumps, could potentially impede the performance of the Emergency Core Cooling System (ECCS). To evaluate transport of material towards the sump and the potential for degradation in performance of the ECCS, Computational Fluid Dynamics (CFD) has been used to predict the volume of material transported to the sump screens [1]. This predicted volume is then used in full scale laboratory tests to determine head loss across the screen under design flow rates. The laboratory sump strainer tests employed a flume facility measuring 14 m by 3 m by 1.5 m tall with a 2.5 m by 3 m by 2 m deep pit at one end, which can accommodate multiple full scale strainer modules. Head loss performance of the modules under different insulation debris loading conditions was evaluated. The internal walls of the flume were adjusted to reproduce prototypical average approach flow velocity and velocity gradients such that the transport of insulation debris to the strainer modules was accurately represented. A three-port isokinetic sampling system was integrated into the downstream piping for measuring debris bypass. This paper will cover the sump screen head loss testing methodology, and the associated integration of the computational results for the source terms.
ALWR PRA Standard Development
Authors: Chapman Jim , Grantom Rick , True Doug
Published: February 2010

In 2007, the American Society of Mechanical Engineers (ASME) Committee on Nuclear Risk Management (CNRM) formed a working group (WG) to develop probabilistic risk assessment (PRA) requirements for advanced light water reactors (ALWRs). Existing PRA Standard language (internal events, fire and external events) was developed for operating plants. The requirements in the existing, and possibly in the planned Standards, are not fully applicable for an ALWR from the design through initial operational phases. Furthermore the WG had a goal to determine if additional requirements were appropriate for design and operational features which might differ from operating plants. Consider for example: • Prior to completion of construction and initial operation, final design and operational information may not be available, and walkdowns of final, as constructed, features will not be possible; • Plant-specific operating experience will not be available; and • ALWRs may have features, such as passive systems and digital instrumentation and control (I&C), which may not be adequately addressed. The initial scope of this project has addressed internal events at power requirements. To identify new or revised requirements, the ASME Standard, Regulatory Guide 1.200 rev. 01, Regulatory Guide 1.206, and Draft Regulatory Guide 1.200 rev. 02 were reviewed in detail. In addition the results of a peer review of an ALWR in the design stage using the ASME Standard were considered. Draft requirements have been developed and are currently in review by the CNRM. The paper describes the approach, results and future actions to modify the combined Standard to address ALWRs.
Controlling and Supervising the Nuclear (Core) Power of a PWR in Relation to Nuclear Safety and Economics
Authors: Wouters P. , Van Rompay W. , Bertels F. , Van Hove W. , Gorleer E. , Vermeulen L. , Hauser E. , Mediate T.
Published: February 2010

Knowing exactly the nuclear core power of a nuclear reactor is one of the most important parameters for the operator; it is vital for safety as well as for economical matters. The secondary calorimetric is the only one where one can pilot on; it is a combination of measured parameters, of which the feedwater (FW) flow towards the steam generators is the most significant one. This feedwater flow can be measured by means of an ultrasonic flow meter, “LEFM CheckPlus™ system” instead of the commonly used venturis or diaphragms. In the Belgian Nuclear Power Plant (NPP) Doel 4, a new ultrasonic “LEFM CheckPlus™” feedwater flow measuring system has been installed in April 2008. The paper describes the consequences of the installation, as the total error on the secondary calorimetric decreases from the previous 1,3% to the current 0,8% with a possibility of further reduction to 0,4%. Additionally, the economical effects of the installation are calculated for a 1000 MWe power plant with venturi meters undergoing fouling. For the NPP Doel 4 it was an economically interesting investment since the payback period was only 45 days. Finally, the possibility of consuming the margin on the secondary calorimetric for a mini-power uprate is inspected, technically and economically. It is concluded that such a mini-power uprate is an interesting option for the NPP owner.
Competitiveness of Small-Medium, New Generation Reactors: A Comparative Study on Decommissioning
Authors: Locatelli Giorgio , Mancini Mauro
Published: February 2010

Smaller size reactors are going to be an important player of the worldwide nuclear renaissance. The economy of scale plays against the development of this kind of reactors, even if sometimes its influence is overestimated so that Small Medium Reactor (SMR) appears to have a Levelised Unit Electricity Cost (LUEC) significantly higher than Large Reactors (LR). However, the economy of scale applies only if the designs of SMR are similar to that of LR, but this is not the case, since the small size allows original design solutions not accessible to large size reactors. The literature already presents studies showing how, under certain assumptions, the Capital Cost and the Operation and Maintenance (O&M) Cost of a site provided by one large reactor is quite similar to another site composed by 4 SMRs providing the same power. However the literature still lacks of this kind of analysis about the decommissioning cost. The paper fulfill this gap investigating the cost breakdown of a decommissioning project and providing a literature review about its cost estimate techniques and managerial approach.
A Preliminary Approach to the Dynamic Behaviour of Circulating Fuel Reactors for Generation IV
Authors: Di Marcello Valentino , Cammi Antonio , Luzzi Lelio
Published: February 2010

A preliminary approach to the simulation of the dynamics of the Circulating Fuel Reactors (CFR) is presented in this paper. Among CFRs, the most promising is the Molten Salt Reactor (MSR), which is one of the “Generation IV International Forum” concepts. Assuming the space-independent zero-power kinetics equations modified in order to take into account the coupling with the molten salt circulation, the dynamic behaviour has been analyzed by means of the transfer function of the system (core + external primary loop). A very simple model is proposed that has been validated on the basis of the experimental data from the Molten Salt Reactor Experiment (MSRE) performed at the Oak Ridge National Laboratory (ORNL); after the validation, it has been applied to study the effects of the coolant velocity on the reactor stability, taking into account the circulation time in the core as well as in the external primary loop. Even if the material properties adopted in the analyses refer to the values encountered in the MSRE, the dynamic model is of more general validity for nuclear reactors with circulating fuels. The results of the present work allow a preliminary evaluation of the coupling between the point neutron kinetics and the velocity of the circulating fuel, and they also give a first indication on the control strategy to be adopted in the current development of MSRs for Generation IV.
Enhanced VVER-1000 Fuel Technology and Performance
Authors: Shah H. , Latorre R. , Raspopin G. , Sparrow J.
Published: February 2010

The United States Department of Energy, through the Pacific Northwest National Laboratory (PNNL), provides management and technical support for the International Nuclear Safety Program (INSP) to improve the safety level of VVER-1000 nuclear power plants in Central and Eastern Europe.
Continuous-Operation Test at Engineering Scale Uranium Crystallizer System
Authors: Washiya Tadahiro , Tayama Toshimitsu , Nakamura Kazuhito , Yano Kimihiko , Shibata Atsuhiro , Nomura Kazunori , Chikazawa Takahiro , Nagata Masanobu , Kikuchi Toshiaki
Published: February 2010

Uranium crystallization based on solubility difference is one of the remarkable technologies which can provide simple process to separate uranium in nitric acid solution since the process is mainly controlled by temperature and concentration of solute ions. Japan Atomic Energy Agency (JAEA) and Mitsubishi Materials Corporation (MMC) are developing the crystallization process for elemental technology of FBR fuel reprocessing.[1–3] The uranium (U) crystallization process is a key technology for New Extraction System for TRU Recovery (NEXT) process that was evaluated as the most promising process for future FBR reprocessing.[4–6] We had developed an innovative crystallizer and carried out several fundamental investigations. On the basis of the results, we fabricated an engineering-scale crystallizer and have carried out continuous operation test to investigate the stability of the equipment at steady and non-steady state conditions by using depleted uranium. As for simulating typical failure events in the crystallizer, crystal accumulation and crystal blockage were occurred intentionally, and monitoring method and resume procedure were tried and selected in this work. As the test results, no significant phenomenon was observed in the steady state test. And in the non-steady state test, process fluctuation could be detected by monitoring of screw torque and liquid level in the crystallizer, and all failure events are proven to be recovered by appropriate resumed procedures.
RCS Pressure Predictions Under ATWS for Maanshan With MUR
Authors: Wu Wen-Hsiung , Shih Chunkuan , Wang Jong-Rong
Published: February 2010

The reactor coolant system (RCS) pressure transients under ATWS for Maanshan PWR with MUR conditions were analyzed in this study. The main concern was to observe whether the peak RCS pressure still maintained within 3200 psig limit (defined by ASME Code Level C service limit criterion under 100% rated power). The Maanshan RETRAN model, which had been developed for PWR transients, was used for ATWS safety analysis and related sensitivity analysis under MUR conditions. In document SECY-83-293, all initializing events were classified as either turbine trip or non-turbine trip events and their ATWS risks were also evaluated according to these two events. Loss of condenser vacuum (LOCV) and Loss of normal feedwater (LONF) ATWS were identified as limiting transients of turbine trip and non-turbine trip events in this study. In order to analyze the RCS pressure variations, a moderator density coefficient (MDC) model was included to observe the influences of moderator temperature and system pressure to the reactor reactivity, and the possibility of the peak RCS pressure reaching its limit. In addition, this report also covers the following sensitivity studies, (1) different MTC settings, and (2) different initial conditions of steam generator outlet pressure. The peak RCS pressure was found to stay within the criterion of 3200 psig even at a conservative MTC setting of −4 pcm/°F. Furthermore, a lower steam generator outlet pressure also resulted in a lower peak RCS pressure.
Application of Virtual Reality in Evaluation and Optimization for Human Machine Interface
Authors: Yan Shengyuan , Xia Chunyan , Zhang Zhijian , Peng Minjun , Yang Ming
Published: February 2010

Functional simulation based evaluation plays a very important role in human machine interface (HMI) evaluation. Virtual OM system is newly emerged as effective tool for realizing the interaction process of HMI. In this paper, a 3D virtual environment, dynamic operating and monitoring (OM) modules of HMI of main control room are set up by using GL Studio software. Subjective evaluation models and the evaluation indexes of operating and monitoring interface were proposed. In this study, grey clustering method was used for evaluating the designed virtual HMI. The virtual functional simulation HMI system and the subjective evaluation module are integrated into GL studio software. By this way, a virtual functional simulation HMI evaluation and optimizing system has been realized.
Application of Risk-Informed Methods to Improve Plant Operation at Kozloduy Nuclear Power Plant
Authors: Kichev Emil , Ivanov Ivan , Mancheva Kaliopa , Petrov Yasen , Vladimirova Vesselina , Andre Gerald , Stevenson Paul
Published: February 2010

Refueling outages at the Kozloduy Nuclear Power Plant (KNPP) Units 5 and 6 are used to perform annual repairs and preventive maintenance activities, piping inspections, and test activities. A refueling outage at KNPP typically requires 60 days and occurs on an annual basis. Testing of safety systems at the KNPP Units 5 and 6 is an extensive exercise that results in multiple actuations of all components during each test and a relatively high number of component actuations each year. This results in equipment wear out issues that can lead to considerable component replacement and/or refurbishment. Numerous piping in-service inspections are conducted in locations where there has been no industry or plant-specific indications or failures, leading to unnecessary personnel exposure. KNPP is interested in using risk-informed (RI) approaches to reduce refueling outage length, piping inspections, testing, and exposure. KNPP is a four-loop Voda-Vodyanoi Energetichesky Reaktor (VVER) with a power level of 1000 MWe. Safety systems consist of three trains. The KNPP at-power probabilistic safety assessment (PSA) model includes internal and external events. It addresses the full range of events leading to core damage frequency (CDF) and includes a simplified level 2 model leading to large early release frequency (LERF). The RI approach, as defined in the U.S. Nuclear Regulatory Commission’s (NRC’s) risk-informed (RI) Regulatory Guides (RGs) 1.174, 1.177, and 1.178, was used in this program. The specific approach used for risk-informed in-service inspection (RI-ISI) is based on the Pressurized Water Reactor Owner’s Group methodology. The overall approach for each of the three applications used a multi-step process which included the following: identification of systems to address; identification of alternatives to current maintenance, inspection, and testing practices; a risk assessment of the proposed alternatives; an assessment of the impact of the changes on deterministic considerations; identification of monitoring requirements; and an assessment of the economic benefits. The RI-ISI program also considered the impact of the changes on personnel exposure. The overall approach made extensive use of data assessments, reliability methods, and risk assessments. The results demonstrated that the proposed changes in maintenance, in-service inspection, and testing programs have a small impact on risk, based on CDF and LERF. In addition, the proposed changes provide significant benefits in terms of reduced outage time, in-service inspections, testing requirements, and personnel exposure. The economic analysis demonstrated that changes to the maintenance program provide the largest benefit followed by the changes to the in-service inspection program and then the changes to the testing program.
Treatment of Noise in Experimental Transport Measurements Plots With Discrete Wavelet Transforms
Authors: Panta Pazos Rube´n
Published: February 2010

In this work it is applied the wavelet transform method [2] in order to reduce diverse type of noises of experimental measurement plots in transport theory. First, suppose that a problem is governed by the transport equation for neutral particles, and an unknown perturbation occurs. In this case, the perturbation can be associated to the source, or even to the flux inside the domain X. How is the behavior of the perturbed flux in relation to the flux without the perturbation? For that, we employ the wavelet transform method in order to compress the angular flux considered as a 1D, or n-th dimensional signal ψ. The compression of this signal can be performed up to some a convenient order (that depends of the length of the signal). Now, the transport signal is decomposed as [9, 11]: ψ=〈am|dm|dm−1|dm−2|⋯|d2|d1〉 where ak represents the sub signal of k-th level generated by the low-pass filter associated to the discrete wavelet transform (DWT) chosen, and dk the sub signal of k-th level generated by the high-pass filter associated to the same DWT. It is applied basically the Haar, Daub4 and Coiflet wavelets transforms. Indeed, the sub signal am cumulates the energy, for this work of order 96% of the original signal ψ. A thresholding algorithm provides treatment for the noise, with significant reduction in the compressed signal. Then, it is established a comparison with a base of data in order to identify the perturbed signal. After the identification, it is recomposed the signal applying the inverse DWT. Many assumptions can be established: the rate signal-to-noise is properly high, the base of data must contain so many perturbed signals all with the same level of compression. The problem considered is for perturbations in the signal. For measurements the problem is similar, but in this case the unknown perturbations are generated by the apparatus of measurements, problems in experimental techniques, or simply by random noises. With the same above assumptions, the DWT is applied. For the identification, it is used a method evolving statistical and metric techniques. It is given some results obtained with an algebraic computer system.
Economic Analysis Regarding the Introduction of a CANFLEX-NU Fuel Into Wolsong Units
Authors: Park J. H. , Jung J. Y. , Kang D. H. , Moon P. W.
Published: February 2010

A CANDU core is composed of lots of fuel channels of a pressure tube type which creates an independent flow passage different from that of a LWR. Hence, most of the aging effects for a CANDU’s operating performance originate from a crept pressure tube as its operating years proceed. Since the early 1990’s, Korea Atomic Energy Research Institute (KAERI) and Atomic Energy of Canada Limited (AECL) have cooperated to develop, verify, and demonstrate an advanced CANDU fuel, called a CANFLEX-NU (CANdu FLEXible-Natural Uranium) fuel, which has been specially designed so as to compensate for a deterioration of the heat transfer rate in a crept pressure tube and also to enhance the safety margin of it by a reduction of the maximum linear element power of a fuel bundle. Now, a CANFLEX-NU fuel is ready to be commercialized in a CANDU-6 because its design and demonstration irradiation have been completed in both of Korea and Canada. Recently, some CANDU plants are being refurbished as they are approaching their life time. One of the refurbished CANDU’s is Wolsong Unit 1 in Korea. It has been operated for 25 years and the operating power at the present time is less than 90% of a full power because of a reduction of the margin for the ROP trip set point. The most appropriate way, up-to now, to overcome such a power de-rating due to a crept pressure tube is the introduction of a CANFLEX-NU fuel into a CANDU. Before commercializing a CANFLEX-NU fuel in the Wolsong Units, it was required to estimate the economic benefit analysis for a CANFLEX-NU fuel in the Wolsong Units because the Uranium price at present is very different from that during the development phase of the CANFLEX-NU fuel. The present paper proposes an economic evaluation model for a CANFLEX-NU fuel in the Wolsong Units and its result for a CANFLEX-NU fuel are compared with that for a 37-element fuel. It was found that the introduction of a CANFLEX-NU fuel into the Wolsong units would have economical benefits due to a better operating performance even where it is assumed that the fuel fabrication cost of a CANFLEX-NU fuel is 1.4 times that of a 37-element fuel. Once the final results are obtained, they will be used as one of important parameters to determine the introduction of a CANFLEX-NU fuel in the Wolsong Units, or not.
Experimental Evaluation of Wire Type Leak Detector Layout for Prototype Fast Breeder Reactor (PFBR)
Authors: Rajan K. K. , Vijayakumar G. , Chandramouli S. , Madhusoodhanan K. , Kalyanasundaram P. , Vaidyanathan G.
Published: February 2010

Wire type leak detectors working on conductivity principle are used for detecting sodium leak in the secondary sodium circuits of FBRs. It is required to assess the performance of these detectors and confirm that they are meeting the requirements. A test facility by name LEENA was constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam to test the wire type leak detector lay out by simulating sodium leaks of different rates. This test facility consists of a sodium dump tank, a test vessel, interconnecting pipelines with valves, micro filter and test section with leak simulators. There are three different test sections in the test set up of length 1000 mm each. These test sections simulate piping of Prototype Fast Breeder Reactor (PFBR) secondary circuit and the leak detector layout in full scale. All test sections are provided with leak simulator. A leak simulator consists of a hole of size one mm drilled in the test section and closed with a tapered pin. The pin position is adjusted by a screw mechanism and there by the annular gap of flow area is varied for getting different leak rates. Test facility was commissioned and 20 experiments were attempted at 350°C to 550°C. Out of 20 experiments 11 experiments were successfully completed and 9 experiments were terminated in between due to the choke in the simulator hole. From the experimental data it is found that sodium leak rate of 200 g/h and above can be detected within 6 hours. A relationship between leak rate and detection time was established from the experimental results and found that sodium leak rate of 100g/h is likely to be detected in 11.4 hours. This paper deals with the details of wire type leak detector layout for the secondary sodium circuit of PFBR, performance requirement of leak detection system as per codes, description of test facility, experimental procedure and test results. Paper also reviews the experiment conducted in CEA, Cadrache and compares with results of present experimental study.
Sustainable Development of China With Nuclear Energy System Beyond Generation-IV
Authors: Zhou Zhiwei , Xu Hong , Yang Yongwei
Published: February 2010

Two aspects of the development trend of current nuclear fission technology are discussed. The first aspect is to improve economic competitiveness and safety for searching opportunity of enlarging the share of nuclear power. The second aspect is to explore new ways of improving the efficiency of nuclear fuel utilization and of reducing the geological repository volume of radioactive products from nuclear power generation. Sustainable development of Chinese economy in 21st century will mainly rely on sustainable supply of clean energy with indigenous natural resources. The burden of current coal-dominant energy mix and the environmental pollution due to energy consumptions has led nuclear power to be an indispensable choice for further expanding electricity generation capacity and for reducing greenhouse effect gases emission in China. The long-term sustainable development strategy with nuclear fission technology beyond generation-IV for electric power generation, namely the fusion-fission hybrid subcritical reactor technology, is discussed. The impact of the proposed fission-fusion hybrid reactor to future nuclear power generation technology will reply on the success of the ITER-scale (500MW fusion power) Tokamak to burn plasma continuously in the predictable future. The main challenges and prospects of the strategy are also analyzed. The preliminary analysis has shown that the fission in the subcritical blanket driven by fusion neutrons can effectively amplify the energy carried by fusion neutron and maintain breeding of fissile material and tritium. It has been found from the results of a conceptual design that this new type of fusion-fission hybrid reactor may meet the requirement of China’s long-term sustainable development of nuclear energy.
Reaction Behavior of Carbon Dioxide With Liquid Sodium Pool
Authors: Miyahara Shinya , Ishikawa Hiroyasu , Yoshizawa Yoshio
Published: February 2010

Reaction behavior of carbon dioxide (CO2) with a liquid sodium pool was experimentally investigated to understand the consequences of boundary tube failure in a sodium-CO2 heat exchanger. In this study, two kinds of experiments were carried out to investigate the reaction behavior. In one experiment, about 1–5g of liquid sodium pool were poured into flowing CO2 to obtain the information mainly about the thermo-chemical conditions to initiate the reaction and the chemical constituents of reaction products. During the experiment, visual observation was made using video-camera and the temperature change of the sodium pool and near the surface was measured by thermocouples. The experimental parameters were the sodium pool diameter, the initial temperature of sodium and CO2, the CO2 flow direction against pool surface, and the initial moisture concentration in CO2. The solid products of sodium-CO2 reaction were sampled and analyzed by X-ray diffraction (XRD), Energy Dispersion X-ray analysis (EDX), Total Organic Carbon analysis (TOC), and chemical analysis. The reaction gas products were also sampled and analyzed by gas chromatography. In the other experiment, CO2 was injected into about 200g of liquid sodium pool to simulate the boundary failure in the sodium-CO2 heat exchanger. The CO2 was fed through a helical coil-type tube dipped into the pool to adjust the temperature to the sodium pool temperature, and injected upward into the pool from a pool bottom using a nozzle attached at the end-side of the tube. The experimental parameters were the initial temperature of sodium, the diameter of the nozzle, the flow rate and the injection time of CO2. The temperature change of sodium pool and the cover gas was measured by thermocouples during the experiment, and the reaction products were sampled and analyzed by the same manner as in the former experiments after the experiment. From these experiments, it became clear that the exothermic reaction occurred above a threshold temperature, and useful and indispensable information such as the resulting temperature and pressure rise and the behavior of solid reaction products in the pool was obtained to evaluate the consequences of boundary tube failure incident in a sodium-CO2 heat exchanger.
Derivation and Analysis of the Criterion for Onset of Density Wave Oscillation in Tubes and Channels With Riser
Authors: Kornienko Yu.
Published: February 2010

In a number of nuclear power plant (NPP) applications questions arise as to appearance and development of oscillations in mass flow rate, enthalpy, coolant density in the channels with riser. Density waves oscillations (DWO) are one of the most common kind of self-oscillating processes in the parallel steam-generating channels system. It concerns inter-channel pulsations as well as oscillations in natural convection loops with a periodical space-time density change propagating together with the flow. That is why obtaining an analytical criterion giving the boundary for onset of DWO is of the notable theoretical and practical interest.
The Importance and Meaning of Signatures and Integrity to the Global Nuclear Community and Successful Nuclear Renaissance
Authors: Smith Clayton T.
Published: February 2010

Welcome to the “Nuclear Renaissance”... The global nuclear renaissance and concerns over climate change were addressed by the keynote speakers at the 20th anniversary of POWER-GEN International, December 2, 2008. J.M. Bernhardt, Jr., chairman, president and CEO of Shaw Group, opened his presentation by acknowledging that the nuclear renaissance is not something that might happen; it is happening now. He stated that “When greenhouse gasses are in the mix, he believed nuclear is where we need to go.” Additionally, Jacques Besnainou, president and CEO of Areva, Inc supported Bernhardt with the call for more investment in nuclear power. He stated that “nuclear power is not THE solution, but there is no solution without nuclear power. What Will it Take to Make the Global Nuclear Renaissance a Success?Must have the need... The Department of Energy (DOE) indicates U.S. energy consumption is expected to grow 1.5 percent a year over the next two decades, and should reach an increased 45% baseload need by 2030. Additionally, the gap between total worldwide energy production and demand will double over the next 15–20 years. Must have the will... Nuclear Energy Institute (NEI), and World Nuclear Association indicate that globally, 36 plants are currently under construction, 93 plants are on order or planned in 18 countries (30 in U.S.), and 219 projects are under consideration in 27 countries. Additionally, they indicate that the combination of low production costs, high reliability, safe operation and clean air benefits positions nuclear energy favorably to meet our baseload electricity needs today and for future expansion. Must have the commitment... Each of us who are members of the global nuclear community must have our core values embedded in a solid quality foundation. We exercise our commitment to quality daily by the application of our signature on documents, instructions, and drawings.
SAIPEM Novel Hybrid Riser: RCS
Authors: Casola Floriano , Mahoney Garry , De Amicis Luca
Published: February 2010

As the offshore industry continues to progress developments in deep and ultra-deep water locations, in many large EPCI contracts (Engineering, Procurement, construction and Installation) the considered or preferred riser system solutions are steel catenary risers (SCRs) or lazy wave (LW) SCRs. The criticality of SCR technology continues to require detailed assessment of technology risks for each specific development, even where design conditions may not appear to vary significantly from previous developments. The potential for increased water depth, severity/complexity of environmental loading, a wider range of vessel typology and with ongoing integrity monitoring & maintenance feedback, it is clear that there is an increasing risk that SCR will exceed the technological limits for near future developments. Looking forward, there is a primary need to manage the risks of this elegant & commercially attractive riser solution by ensuring the margin between the expected fatigue /extreme loading and the current technological limits of metallic material/welding, top connection and non-metallic components. In this context, and through a fully integrated approach to riser technology, SAIPEM has conceived and matured families of hybrid riser systems based on field proven technology. One of these families is the novel hybrid riser family, specifically conceived to extend the application of SCR & LWSCR in a cost effective way as well as offering certain additional improvements. The basic solution is the RCS (Novel Hybrid Riser with Top Connecting Arm & Shallow Water Jumper); a balanced cost /risk hybrid riser solution to compete with SCR & LWSCR. RCS offers consistent improvements and is adaptable to the major vessel typologies, from semi-subs to disconnectable turret moored FPSOs. The RCS comprises Top Connecting Arm (TCA), Riser Top Assembly (RTA) and Tension Restrained Jumper (TRJ) or simple short/long catenary jumper (SCJ), and is connected to the vessel by means of a deepwater type mooring connector. RCS+ is a further evolution of RCS incorporating adjustable length TCA. RCS+ Heave Motion Damper (HMD) is a future evolution of RCS, and is described separately. The RCS is an example of the application of an industrial model that includes an integrated engineering and execution expertise to achieve future cost effective solutions.
Numerical Simulations Concerning the Tendency of Soil Plugging in Open-Ended Steel-Piles
Authors: Henke Sascha , Grabe Ju¨rgen
Published: February 2010

The development of soil plug inside open-ended steel-piles is investigated using Finite-Element method. The penetration of a rigid pile with various diameter into granular soil is simulated numerically to better understand the mechanisms which occur during pile installation inside the open-ended piles. The numerical results are analyzed for a better understanding of the influence of the installation method on soil plugging. The received results are compared to experimental results out of literature. Concluding, a parametric study is fulfilled to examine the role of soil density and pile diameter concerning the tendency of plug formation inside a jacked pile.