bluuebunny/crossref_metadata_2025 huggingface dataset:
Applying Engineering Simulator to Verification and Validation of Digital I&C in Nuclear Power Plant
Authors: Lin Meng , Yang Zongwei , Hou Dong , Liu Pengfei
Published: February 2010

In China, more and more Nuclear Power Plant (NPP) will be constructed in the near future years, and Main Control Room (MCR) will introduce digital Instrumentation and Controls systems (I&C) technique. I&C system of nuclear power plant consists of Control Systems, Reactor Protection System and Engineered Safety Feature (ESF) Actuation System. For example, I&C system of LinAo Phase II NPP has adopted SIEMENS TXP and TXS I&C, which is being constructed in Guangdong province, China. In this engineering project, Chinese engineers are responsible for all the configuration of actual analog and logic diagram. Before the phase of real plant testing on the reactor, engineers want to make sure that configuration is right and control functions can be accomplished, so primary Verification and Validation (V&V) of I&C works were done. One way is checking the diagram configuration one by one according to the original design. There are two main disadvantages. One is diagram is so complex that workload is very large and engineers will make mistake. Another is even engineers have read every logic, but they still cannot know the final results and function of a complex control system. So another effective V&V way is applying NPP engineering simulator to do virtual test. According to LinAo Phase II NPP design, we develop one simulator to construct a virtual NPP model as a basis, which can provide plant operation parameters and can also accept control signal from I&C, then give response to it. Through this way, we don’t need to know the exact diagram, and just observe input and output of I&C to make sure that the final results is right and functions have been accomplished. In this way, it is need to transfer signals between simulator and I&C. For keeping the original software and hardware structures of SIEMENS Distributed Control System (DCS), we use one set of data acquisition (DAQ) equipments to build a connection between the engineering simulator (software) and SIEMENS DCS I/O cabinet (hardware), and the interface is standard 4–20mA direct current and 0–48V direct voltage. This way is convenient for expansion to other digital I&C V&V. After these two V&V works, we can then build the confidence of digital I&C control function. As an application research, we mainly focus on V&V of digitalized control systems and selected several Reactor Control (RRC) systems as examples, including pressurizer pressure and water level control, steam generator water level control. In this paper, we will introduce the way of applying engineering simulator to do V&V works, the structure of our simulator, the function of different block, and primary V&V results. Moreover, we will have ideas on the future application of this methodology to V&V of Reactor Protection System and ESF Actuation System.
On-Line Monitoring the In-Core Flux Distribution Using the Eigenfunctions
Authors: Wang Changhui , Wu Hongchun , Cao Liangzhi
Published: February 2010

A method of on-line monitoring the reactor core flux distribution is proposed in this paper. In this method, the real in-core flux distribution is reconstructed on-line using the eigenfunctions of neutron diffusion equation. we expand the real in-core flux distribution by the eigenfunctions with linear combination and calculate the expansion coefficients according to the fixed in-core detector readings which are the real-time response of in-core flux distribution. For only a few expansion coefficients need to be calculated in monitoring process, this method has a high calculation speed to meet the requirement of on-line monitoring the reactor. A program named NLSANM-K (Nonlinear Iteration Semi-Analytic Nodal Method & Krylov subspace method) is encoded to calculate the eigenfunctions. The accuracy of the reconstruction is verified by comparing the reconstructed flux distribution with the results calculated by a core simulator SAMME.
Study on Properties of Reprocessed Uranium: Investigation of Surface Gamma Dose and Water Content of Reprocessed UO3 Powder in the Storage
Authors: Yamanaka Atsushi , Murakami Manabu , Goto Yuuichi , Senoo Shigeo
Published: February 2010

Very small amount of U-232 is included in UO3 powder reprocessed from spent fuel. Bi-212 and Tl-208 (those are) the daughter nuclides of U-232 are the major gamma sources and they increase the gamma dose rate on the surface of UO3 storage container. So, it is necessary to pay attention to the exposure doses of workers. In fact, the surface gamma dose of the container is increasing in Tokai Reprocessing Plant of JAEA. Therefore, UO3 samples were taken from the storage containers and the properties were investigated. Relation between the Bi-212 and Tl-208 contents and the surface gamma dose of the storage container was clearly observed, however their calculated doses were 30–50% of the measured ones. Water content of UO3 powder was also investigated and it was kept 0.2wt% or less for over 20 years.
Residual Life Assessment of Steam Generators With Alloy 600TT Tubing: Methods and Application
Authors: de Curieres Ian , Meunier Marie-Christine , Joly Pierre
Published: February 2010

The aim of the paper is to introduce methods to estimate the residual life of steam generators with alloy 600TT (thermally treated) tubing, taking into account PWSCC as the main contributor damage. The methods take into account both initiation and propagation of PWSCC cracks in the expansion transition zone of SG tubes, as well as the state of damage (cracking and plugging) of the tube bundle, at a given time, known from inspection results. A probabilistic model is used to treat initiation, while propagation stage is treated in a deterministic way based on inspection data. After introducing the methods used to assess the residual life, a brief parametric study will be shown, to illustrate the effects of initiation versus propagation. Eventually, the cases of a few actual steam generators with tubing made of alloy 600TT, showing different situations of present damage and damage evolution rates will be presented.
A Comparative Study of Single-Phase, Two-Phase and Supercritical Natural Circulation in a Rectangular Loop
Authors: Vijayan P. K. , Pilkhwal D. S. , Sharma M. , Saha D. , Sinha R. K.
Published: February 2010

A one dimensional theoretical model has been used to analyze the steady state and stability performance of single-phase, two-phase and supercritical natural circulation in a uniform diameter rectangular loop. Parametric influences of diameter, inlet temperature and system pressure on the steady state and stability performance has been studied. In the single-phase liquid filled region, the flow rate is found to increase monotonically with power. On the other hand the flow rate in two-phase NCS is found to initially increase, reach a peak and then decrease with power. For the supercritical region also, the steady state behaviour is found to be similar to that of two-phase region. However, if the heater inlet temperature is beyond the pseudo critical value, then the performance is similar to single-phase loops. Also, the supercritical natural circulation flow rate decreases drastically during this condition. With increase in loop diameter, the flow rate is found to enhance for all the three regions of operation. Pressure has a significant influence on flow rate in two-phase region marginal effect in supercritical region and practically no effect in the single-phase region. With increase in loop diameter, operation in the single-phase and supercritical regions is found to destabilize whereas the two-phase loops are found to stabilize. Again, pressure has a significant influence on stability in the two-phase region.
Experimental Investigation of the Influence of Flow Obstacles on Post-Dryout Heat Transfer in an Annulus
Authors: Anghel I. G. , Anglart H. , Hedberg S. , Rydstro¨m S.
Published: February 2010

This paper describes the experimental setup, instrumentation and procedures which have been developed in the thermal-hydraulic laboratory at the Royal Institute of Technology (KTH), Stockholm, Sweden, to perform new post-dryout heat transfer investigations in an annulus with flow obstacles. Previous investigations performed in the same laboratory indicated that flow obstacles had a considerable influence on the post-CHF heat transfer. The measured heat transfer enhancement was significantly under-predicted by existing models. However, the net effect of obstacles could not be deduced from the measurements, since reference - obstacle-free measurements - had not been performed. In addition, the number of thermocouples that could be installed inside the heated rod was limited to 8. These deficiencies have been removed in the current approach. Firstly, the present design of the test section allows for measurements both with and without flow obstacles. In this way the net effect of the obstacles will be captured. Secondly, a newly developed technique allowed the installation of 40 thermocouples inside of the heated rod. An additional 40 thermocouples have been installed on the external wall of the heated tube. Therefore, a significant improvement of the accuracy of measurements can be expected. The present arrangement of instrumentation is suitable to perform measurements of heat transfer under both steady-state and transient conditions.
Chromatography Column System With Controlled Flow and Temperature for Engineering Scale Application
Authors: Watanabe Sou , Goto Ichiro , Sano Yuichi , Koma Yoshikazu
Published: February 2010

Japan Atomic Energy Agency (JAEA) is conducting R&D of the engineering scale extraction chromatography system, which uses silica-based adsorbents impregnated with an extractant for the minor actinides (Am and Cm) recovery from the high level liquid waste generated in the spent FBR fuel reprocessing, as a part of the Fast Reactor Cycle Technology Development (FaCT) project. A bench scale testing system was made and provided for the first step of development. The column of the test system (ID 480 or 200 mmΦ with 650 mm height) was equipped with ports for 6 external sensors at its top, middle and bottom levels for measuring the flow velocity or temperature, and for additional 6 heaters for simulating the decay heat of Am and Cm at the middle level of the column. The flow velocity distribution was almost constant except for the very near at the column wall, and it was almost uniform when the liquid flew from top to bottom direction with 4 cm/min of the velocity. The heaters scarcely influenced on the temperature profile inside the column when the power applied to the heater simulated the decay heat of Am, Cm and FPs. The decay heat generated in the column was transported to the effluents and the temperature inside column was kept almost constant.
Characteristics of the New Embrittlement Correlation Method for the Japanese Reactor Pressure Vessel Steels
Authors: Soneda Naoki , Nomoto Akiyoshi
Published: February 2010

Neutron irradiation embrittlement of reactor pressure vessel steels is an important ageing issue for the long term operation of light water reactors. A new embrittlement correlation method was developed by CRIEPI and the Japanese electric utilities in 2007. This method is primarily based on the fundamental understandings on the embrittlement mechanisms: i.e. microstructural changes were modeled by the mathematical form of rate equations, and the predicted microstructural changes were further correlated with the mechanical property changes in transition temperature region. The coefficients of the rate equations were optimized using the Japanese surveillance data of RPV embrittlement. This method was adopted as the revision of the Japanese code, JEAC 4201–2007, in 2007. In this paper, after a brief explanation on the new correlation method, the predictions of the new method will be investigated through comparisons with the previous correlation, JEAC4201–2004, and the US surveillance data in order to identify the characteristics of the new method.
Comparative Analysis of Star-CCM, Star-CD, and CFX Codes for Free-Surface Flow Simulation of the Windowless XT-ADS Spallation Target Water Experiment
Authors: Batta A. , Class A. G.
Published: February 2010

Accelerator driven systems for transmutation of nuclear waste require a coupling of an accelerator to a sub-critical reactor core by means of a spallation target. A windowless target is currently developed in the Integrated Project EUROTRANS of the European Community. In the target, HLM (heavy liquid metal) flows downwards through a concentric feeder surrounding the beam tube and forming a conical free surface which is subjected to a beam current. This results in a very high heat generation. The thermal-hydraulics of the target must ensure that the heat deposited into the HLM can be safely removed. Simulations using a volume of fluid method accounting for mass transfer across the free surface can predict the shape of the conical surface, flow detachment, and transient behavior. In the codes used in this study the mass transfer is accomplished via cavitation models, so that evaporation/condensation occurs at a prescribed vapor pressure. Previous simulations have lead to the design rules for the target. In particular, the velocity in feeder nozzle is limited to 2.5 m/s. Forced detachment by enlarging the guide tube stabilizes the free surface. Cavitation in the feeder nozzle is suppressed by fins which induce the required pressure loss. Recirculation flow in the thermally highly loaded zones is minimized by a small injection angle at the feeder nozzle outflow. The simulation results lead to the design of a water experiment, which precedes the test of a HLM target. In the paper the proposed target geometry of the water experiment is presented. Code-to-code comparisons show qualitative agreement between 3 different CFD codes. The considered codes, Star-CD, Star-CCM, and CFX, yield a stable free surface configuration with small recirculation, necessary for safe heat removal. However, differences are observable with respect to the sharpness of the free surface representation and also the behavior near the axis. Also strong differences in the required computational time exist between different codes. Various structured computational meshes are studied. In some simulations a fine mesh is employed within the feeder only and a substantially coarser grid is used in the free surface region and yet yields results comparable to an all-over fine-mesh simulation.
Newly Developed Two-Tip Optical-Fiber Probe for Accurate Measurement of High-Speed Micro-Droplets: Simultaneous Measurement of Their Velocities and Diameters
Authors: Matsuda Keisuke , Ozawa Yusuke , Saito Takayuki
Published: February 2010

Optical fiber probing is very useful and reliable for bubbles/droplets measurement particularly in the gas-liquid two-phase flows that have dense dispersed phase and are impossible to be measured via usual visualization techniques. For the practical purpose of small- or medium-size bubbles/droplets measurement, one of the authors successfully developed a Four-Tip Optical-fiber Probe (F-TOP) and reported their excellent performance in industrial uses. Recently, particular demands for measuring properties of micro bubbles/droplets have increased in researches on multi-phase flows. However, no one succeeded in simultaneously measuring diameters and velocities of high-speed micro-droplets (velocity > 50 m/s; 50 μm < diameter < 500 μm). We made a challenge of measuring such tiny droplets via newly developed optical fiber probe equipped with two tips (Two-Tip Optical-fiber Probe: T-TOP). We have succeeded in this difficult measurement with it. Each optical fiber probe composing the T-TOP is made of a silica optical fiber (125 μm in external diameter, 50 μm in core diameter, 37.5 μm in clad thickness). The optical fiber was fine-drawn using a micro pipette puller, and this yielded a sub-μm-scale tip. The interval of the fiber axes and the gap of the tips were arranged depending on the droplets diameter range. In this paper, we demonstrate the performance of the T-TOP. First, we confirm its practicality in industrial use. The strength of the T-TOP is confirmed by exposure test of high-velocity and high-temperature steam flows. Second, we consider the influence of the flow on the measurement of T-TOP; the optical noise due to probe vibration by the high-velocity gas flow around the T-TOP is considered. Next, we confirm its performance using an orifice-type nozzle (300 μm < droplets diameter < 500 μm; droplets velocities < 40 m/s). We confirm the performance of the T-TOP; the results of T-TOP are compared with those of the visualization of the droplets by using an ultra-high-speed video camera. At the same time, we consider the process of droplet contact with the T-TOP via visualization of ultra-high-speed video camera.
The Parameters Identification of Time - Delay Models of Heat Transfer Systems From Input-Output Measurements
Authors: Simeunovic´ Goran , Zitek Pavel , Chysky´ Jan
Published: February 2010

The paper presents the identification issues and proposes a parameter identification algorithm that separates the system parameters from the time-delays for a class of single input single output (SISO) linear time delay systems (LTDS). The presence of the unknown time delay greatly complicates the parameters estimation problem, because the parameters of the model are not linear with respect to the time-delay. However, once the time delay is determined, the model becomes linear for the other parameters and hence the common least square method can be utilized directly. Motivated by the nonlinear least squares problem developed in the paper Golub and Pereyra (1973), a novel modification of the so-called variable projection functional is worked out for identification of time delays. In this way, the parameters estimation is separated from the estimation of time delays and the large errors in the parameter estimates, in the case of presence of errors in the time-delay identified values, are avoided. Namely, the small error in the time-delay identified values may often cause a large error in the system parameters identification. A hybrid optimalization method combining a Genetic Algorithm and Nelder-Mead technique is used for minimization of variable projection functional for the identification of time delays. This approach is illustrated by a particular application in the field of heat transfer, concretely on the time-delay model of the recuperative heat exchanger.
Bayesian Analysis of Piping Failure Frequency Using OECD/NEA Data
Authors: Wang M. , Pandey M. D. , Riznic J.
Published: February 2010

The estimation of piping failure frequency is an important task to support the probabilistic risk analysis and risk-informed in-service inspection of nuclear power plant systems (NPPs). Although various probabilistic models have been proposed in the literature, this paper describes a hierarchical or two-stage Poisson-gamma Bayesian procedure to analyze this problem. In the first stage, a generic distribution of failure rate is developed based on the failure observations from a group of similar plants. This distribution represents the interplant (plant-to-plant) variability arising from differences in construction, operation and maintenance conditions. In the second stage, the generic prior obtained from the first stage is updated by using the data specific to a particular plant, and thus a posterior distribution of plan specific failure rate is derived. The two-stage Bayesian procedure is able to incorporate different levels of variability in a more consistent manner. The proposed approach is applied to estimate the failure frequency using the OECD/NEA pipe leakage data for the U.S. nuclear plants.
Operating Experience With Mechanical Seals in Reactor Shutdown Cooling Pumps
Authors: Staniewski G. , Keshavarz A. , Ferguson W. J.
Published: February 2010

The CANDU nuclear reactor has a dedicated shutdown cooling system (SDC) to remove heat from the primary heat transport system and transfer it to the high pressure service water open system. There are four individual SDC circuits, each connected to one of the four quadrants of the reactor. The four circuits are independent and each of them consists of one SDC pump and one heat exchanger. When the reactor is within normal operation, all four SDC loops are isolated but kept pressurized to the reactor inlet header of 9.7 MPa (1410 psi) and warmed up to 93°C. During reactor shutdown condition and at reactor start-up operation, the SDC pumps are in-service. They operate in several different configurations and are exposed to a number of different pressure conditions. This paper will present typical failures of the seal components.
Use of PRA in the Design of the Westinghouse AP1000 Plant
Authors: Lutz Robert J. , Scobel James H. , Anderson Richard G. , Schulz Terry
Published: February 2010

Probabilistic Risk Assessment (PRA) has been an integral part of the Westinghouse AP1000, and the former AP600, development programs from its inception. The design of the AP1000 plant is based on engineering solutions to reduce or eliminate many of the dominant risk contributors found in the existing generation of Pressurized Water Reactors (PWRs). Additional risk reduction features were identified from insights gained from the AP1000 PRA as it evolved with the design of the plant. These engineered solutions include severe accident prevention features that resulted in a significant reduction in the predicted core damage frequency. Examples include the removal of dependencies on electric power (both offsite power and diesel generators) and cooling water (service water and component cooling water), removal of common cause dependencies by using diverse components on parallel trains and reducing dependence on operator actions for key accident scenarios. Engineered solutions to severe accident consequence mitigation were also used in the AP1000 design based on PRA insights. Examples include in-vessel retention of molten core debris to eliminate the potential for ex-vessel phenomena challenges to containment integrity and passive containment heat removal through the containment shell to eliminate the potential for containment failure due to steam overpressure. Additionally, because the accident prevention and mitigation features of the AP1000 are engineered solutions, the traditional uncertainties associated with the core damage and release frequency are directly addressed.
Preliminary Set Up of the Methodology to Analyze Nuclear Fuel Cycle Scenarios
Authors: Vezzoni Barbara , Forasassi Giuseppe
Published: February 2010

This work focuses on the definition of a suitable methodology to compare different nuclear energy options, with particular attention to strategies able to minimize nuclear wastes. In order to set possible transition scenarios from actual GEN-II reactors, the energy demand and the nuclear energy share up to 2050 have been evaluated by means of data available in the international literature. To compare the different options, major indicators of energy-electricity sector have been analyzed. With particular attention to nuclear energy social acceptability the major indicators investigated are the inventory and the radiotoxicity vs. time of residual wastes. In order to reach a final equilibrium situation, transition scenarios, where GEN-III (e.g. EPR) and GEN-IV new reactors (as lead cooled fast ones), have been analyzed. As cases study Spain and France scenarios have been considered.
Situation of the Baffle-Former Bolts in Belgian Units
Authors: Ge´rard Robert , Somville Fre´de´ric
Published: February 2010

The baffle to former bolts are used in Pressurized Water Reactors to attach the baffle plates to the former plates in the reactor vessel lower internals. The resulting structure forms a boundary for the flow of coolant and provides lateral support to the fuel assemblies. Some edge bolts are also present, assembling together the baffle plates. After an operating time of the order of 120 000 hours, some bolts exhibit cracking at the junction of the head and the shaft of the bolt. Examinations of failed bolts have made it possible to identify the cause of cracking as irradiation assisted stress corrosion cracking (IASCC). Up to now, baffle bolt cracking has been detected in units older than 15 years, where the baffle bolts are not cooled (no holes in the former to allow a water flow on the bolt shaft). In Belgium the concerned unit are Tihange 1 and Doel 1–2. The paper summarizes the experience with baffle bolts cracking in Belgian units and the strategy implemented to mitigate this problem, consisting of structural integrity analyses, baffle bolts inspections and replacement, and research programs in the field of IASCC, including examinations of highly irradiated replaced bolts.
Application of a Step-Wise Verification and Validation Procedure to the 3D Neutron Kinetics Code DYN3D Within the European NURESIM Project
Authors: Rohde U. , Mittag S. , Grundmann U. , Petkov P. , Ha´dek J.
Published: February 2010

A generic strategy of core physics codes benchmarking was elaborated within the European NURESIM code platform development. In this paper, the application of this step-wise procedure to benchmarking the 3D neutron kinetics code DYN3D for applications to VVER-type reactors is described. Numerical and experimental benchmark problems were considered for code verification and validation. Examples of these benchmarks including benchmark set-up and results obtained by use of DYN3D in comparison with other codes are given. First, mathematical problems with given cross sections are used for the verification of the mathematical methods applied e.g. in nodal codes against finite difference solutions. Discretization errors were quantified. After minimisation of numerical errors, modelling errors have to be considered. Diffusion approximation and homogenisation errors are due to simplified physical approaches and can be estimated by comparing diffusion solutions with more accurate Monte Carlo or deterministic transport solutions. Methods to reduce these errors are outlined. A series of 2D whole core benchmarks for different core loadings and operational conditions for VVER-1000 reactors was defined for this purpose. Reference transport solutions were calculated by the MARIKO and APOLLO codes based on Method of Characteristics. Homogenised two-group and few-group diffusion parameters were derived from the reference solutions and used as cross section data for the nodal diffusion code DYN3D. The DYN3D solutions were compared to the reference solution. It was shown that the homogenisation error can be significantly reduced by using Assembly Discontinuity Factors (ADF) and Reflector Discontinuity Factors (RDF) which are obtained from the transport solution by applying equivalence theory. A study using the multi-group version of DYN3D has shown that increasing the number of groups in the considered cases has only a small effect in comparison with homogenisation error. After reducing modelling errors by choosing appropriate physical approximations, the code have to be validated against reality. Experimental problems are used for code validation. Experimental data for VVER reactors, which were used for the benchmarking of the DYN3D code within NURESIM, are power distribution measurements at the full-size (VVER-1000) experimental facility V-1000, which have been well documented within the EC project VALCO, and kinetic experiments at the LR-0 zero power reactor in NRI Rˇezˇ. The code DYN3D, being one of the NURESIM platform codes, has proved to be an effective tool for steady-state and kinetics core calculations. The high accuracy of the code is based on the advanced nodal method “HEXNEM2”, multi-group approach, applying discontinuity factors, and intra-nodal flux reconstruction.
Thermal Evaluation for a Ultra High Temperature Reactor With Pebble Type Fuels
Authors: Fumizawa Motoo , Kosuge Yuta , Horiuchi Hidenori
Published: February 2010

This study presents a predictive thermal-hydraulic analysis with packed spheres in a nuclear gas-cooled reactor core. The predictive analysis considering the effects of high power density and the some porosity value were applied as a design condition for an Ultra High Temperature Reactor (UHTR). The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 °C was achieved in the case of an UHTREX at LASL, which was a small scale UHTR using hollow-rod as a fuel element. In the present study, the fuel was changed to a pebble type, a porous media. Several calculation based on HTGR-GT300 through GT600 were 4.8 w/cm3 through 9.6 w/cm3 respectively. As a result, the relation between the fuel temperature and the power density was obtained under the different system pressure and coolant outlet temperature. Finally, available design conditions are selected.
Concrete Shrinkage Taken Into Account as Crack Width Assessment
Authors: Gallitre Etienne , Naze Pierre Alain , Labbe Pierre
Published: February 2010

In very large buildings with connected walls, such as large Nuclear Power Plants, concrete shrinkage strains have to be considered because of elements differential strains, as required in the new European construction code (EC 2).. The common engineering method consists in considering shrinkage as an equivalent thermal strain, which is in fact computed as internal forces. In EPR, for Flamanville 3 conditions, this method could lead to unusual in the lower part of the buildings, resulting in very high reinforcement ratio hard to set in place, so EDF with its partners have proposed a new methodology based on crack width assessment. At first, we compute the shrinkage differential strains, “εs”, depending on moisture conditions and elements thickness. Then we fix a reinforcement section as a calculation hypothesis in order to estimate the distance “Ss” between cracks, which is independent from the loads. Consequently, we can assume that a certain crack width value “Ws” is consumed by the shrinkage itself, with Ws = εs Ss Rax (Rax restriction factor). So the available crack width for the other loads is the remaining crack width.. From cracking theory and according to EC2; we can deduce steel stress σd. So the structure design (reinforcement mainly) can be undertaken with this allowable limit value σd., in the load combinations where shrinkage as to be considered. This new methodology is more realistic than the equivalent thermal load based method, so safety requirements are satisfied by focusing on crack width assessment: this is a performance approach. It has allowed EDF to size a reinforcement ratio, which is compatible with concrete technical rules, especially in the areas near the raft. Of course this method remains in accordance with durability hypotheses and other requirements connected to nuclear specificities.
Identification and Prioritization of Generic Nuclear Safety Problems Requiring CFD Analysis
Authors: Smith Brian L.
Published: February 2010

Within the framework of the activities and responsibilities of the OECD/NEA/WGAMA group, a modest PIRT-type exercise was initiated to identify and prioritize country-specific safety issues for which it was considered that analysis using Computational Fluid Dynamics (CFD) could bring real benefits in regard to improved safety. Conventional PIRT procedures were followed as far as possible: the problem scope and objectives of the exercise were first defined, and then a panel of experts was assembled to identify and prioritize the safety items of specific concern to the country represented. Collating the information received, it was possible to identify those safety issues which were clearly country-specific in origin, and those which were of generic interest for which a common approach to benchmarking would be justified. Separate lists were compiled for single-phase and multi-phase phenomena. The paper summarizes the procedures undertaken, lists and prioritizes the safety topics identified in the study, and presents the final ranking in terms of generic importance. Also described are perspectives for a first follow-up activity: the promotion of a new validation benchmark activity based on the problem of high-cycle thermal fatigue in mixing tees.
Nuclear Design Basis Method for Managing Critical Operations Design Content
Authors: August J. K. , Hunter J. J.
Published: February 2010

Designing new plants, clear requirements and efficient processes should translate nuclear designs clearly into operations. Methods that improve translation benefit operations lowering risk, complementing safety and cost performance. Nuclear plant designs develop specifications that assure design integrity from construction into operations throughout plant life. With the certified design at startup, all plants should have (1) scheduled operations monitoring and maintenance plans; (2) developed by qualified engineers; and (3) based on engineering due diligence, equivalent to airline certified operations plans. Although existing rules apply requirements, tort laws have always required operating plans for products.
Development of Ultrasonic Shot Peening Technique for Reactor Components to Improve Structural Integrity Against Stress Corrosion Cracking
Authors: Kawano Shohei , Kawagishi Ayaka , Suezono Nobuichi , Ueno Kenichi , Okuda Ken , Obata Minoru , Kameyama Ikuko , Watanabe Yuusuke
Published: February 2010

The ultrasonic shot peening (USP) technique has been developed for boiling water reactor (BWR) components as a countermeasure against stress corrosion cracking. The effects on residual stress of USP for type 316L stainless steel and alloy 600 were evaluated. Compressive residual stress layer of 0.5 mm from the surface were formed on the specimens after USP using stainless steel ball with a diameter of 3 mm. Cross-sectional hardness measurement revealed that the increase of hardness due to USP is not significant compared with shot peening (SP). The FEM calculation showed the plastic strain induced by the impacts of 3 φ shot with 5 m/s is lower than those of 0.6 φ shot with 50 m/s. It suggests USP process suppresses the degree of work hardening in comparison with SP process. Dissimilar weld joint specimens which simulate the material and dimension of the shroud weld line H7 were examined to confirm the applicability of USP. The experimental result reveals that USP technique is applicable to reactor internal components as stress modification process.
Creation of Automatic System for Monitoring, Forecasting and Control of Condition of Lead-Bismuth (Lead) Coolant and Surfaces of Circuits of Nuclear Power Plants
Authors: Martynov P. N. , Askhadullin R. Sh. , Simakov A. A. , Chaban’ A. Yu. , Chernov M. E. , Lanskikh V. S. , Legkikh A. Yu.
Published: February 2010

Lead-bismuth coolant is preferable for the medium size reactors, since, in contrast to the sodium coolant, it does not interact with water and air, it is radiation resistant, insignificantly activated and it is not combustible [1]. Combination of natural properties of lead-based coolants, mono-nitride fuel, fast reactor neutronics and design approaches used for the reactor core and heat removal system brings SVBR 75/100 NPP [2] to achieve a new safety level and assures its stability without operation of active safety systems even under severe accident conditions. Analysis of possible sequences of the events even under conditions of such severe accidents as addition of total excess reactivity or all pumps trip accompanied by safety system failure leads to the conclusion on that power unit with SVBR 75/100 reactor plant (RP) has high safety level.
Thermal Design Options of New Pressure Channel for SCWRs
Authors: Peiman Wargha , Gabriel Kamiel , Pioro Igor
Published: February 2010

This paper focuses on thermal-design options of a new pressure channel for SuperCritical Water-cooled Reactors (SCWRs). The objectives of this paper are to estimate heat losses from the coolant to the moderator for a preliminary fuel-channel design and to investigate effects of the insulator thickness and moderator pressure on the overall heat losses. In order to fulfill the objectives, the heat losses for an existing reactor were calculated and compared with available values from open literature. These calculations became the basis for calculation of the heat loss for the chosen new fuel-channel design. MATLAB, and NIST REFPROP software were utilized for programming and calculation of thermo-physical properties as needed, respectively. Heat losses for different thicknesses of the ceramic insulator were calculated. These calculations showed that the heat losses for the optimum thickness of insulator, which was calculated to be 7 mm, were about 31 MW. In current CANDU reactors the operating pressure of the moderator is close to the atmospheric pressure; higher operating pressures will allow operation of the moderator at higher temperature while preventing occurrence of boiling in the calandria vessel. Higher moderator temperatures will results in a lower temperature difference between the coolant and the moderator, hence reducing the heat sink from the coolant to the moderator. The effect of the moderator pressure on the heat loss was investigated, which showed that the heat loss can be reduced by increasing the operating pressure of the moderator by approximately 1 MW per 0.1 MPa increase in pressure.
Development and Verification of Unstructured Adaptive Mesh Technique With Edge Compatibility
Authors: Ito Kei , Kunugi Tomoaki , Ohshima Hiroyuki
Published: February 2010

In a design study of the large-sized sodium-cooled fast reactors in Japan (JSFR), one key issue to establish an economically superior design is suppression of a gas entrainment (GE) phenomenon at a free surface in the reactor vessel. However, the GE phenomenon is highly non-linear and too difficult to be evaluated theoretically. Therefore, we are developing a high-precision CFD method to evaluate the GE phenomenon accurately. The CFD method is formulated on an unstructured mesh to establish an accurate modeling for a complicated shape of the JSFR system. As a two-phase flow simulation method, a high-precision volume-of-fluid algorithm is employed in the CFD method. In addition, physically appropriate formulations at gas-liquid interfaces are introduced into the CFD method. The developed CFD method is already applied to the simulation of a GE phenomenon in a basic GE experiment and the simulation results show good agreement with experimental results. Therefore, it is confirmed that the proposed CFD method can reproduce a GE phenomenon. However, for the simulation of the GE phenomenon in the JSFR, we still have one problem on a mesh subdivision. Though a fine mesh subdivision has to be applied to the regions where the GE occurs, it is difficult to preliminarily know the regions because the GE occurrence is strongly affected by a local instant flow pattern, i.e. a vortex generation. Therefore, an adaptive mesh technique is necessary to apply a fine mesh subdivision automatically to only the local GE occurrence regions in the large-sized JSFR. In this study, as one part of an adaptive mesh development, a two-dimensional unstructured adaptive mesh technique is developed and verified. In the proposed two-dimensional adaptive mesh technique, each cell is isotropically subdivided to reduce distortions of the mesh. In addition, a connection cell is formed to eliminate the edge incompatibility between a refined and a non-refined cells. A connection cell has several subdivision patterns and one of them is selected to be compatible with adjacent cells on every cell edge. Finally, the present unstructured adaptive mesh technique is verified by solving well-known driven cavity problem. As the result, the present unstructured adaptive mesh technique succeeds in providing a high-precision solution, although we employ a poor-quality distorted mesh at the initial state. In addition, the simulation error on the unstructured adaptive mesh at the steady state is much less than the error on the structured mesh consisting of a larger number of cells.
Analysis of Severe Accidents in VVER-1000 Reactors Using the Integral Code ASTEC
Authors: Tusheva Polina , Reinke Nils , Altstadt Eberhard , Schaefer Frank , Weiss Frank-Peter , Hurtado Antonio
Published: February 2010

The studies presented are aiming at a detailed investigation of the behaviour of a VVER-1000/V-320 reactor and the containment structures during a postulated severe accident, including the ways and means by which these accidents may be prevented or mitigated. A hypothetical station blackout scenario (loss of the offsite electric power system concurrent with a turbine trip and unavailability of the emergency AC power system), belonging to the typical beyond design basis accidents, has been investigated. Station blackout results in reactor shut down, loss of feed water and trip of all reactor coolant pumps. Continuous evaporation of the secondary side leads to steam generators’ depletion followed by heating up of the core. In case of unavailability of essential safety systems the core will be severely damaged and finally the reactor pressure vessel (RPV) might fail. The analyses are performed using the integral code ASTEC commonly developed by IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and GRS (Gesellschaft fu¨r Anlagen- und Reaktorsicherheit mbH). Code-to-code comparative analyses for the early thermal-hydraulic phase have been performed with the GRS code ATHLET. A large number of sensitivity calculations have been done regarding the axial core power distribution, heat losses, and RPV lower head modelling. The analyses have shown that, despite the considerable differences in the codes themselves, the calculation results are similar in terms of thermal hydraulic response. There are discrepancies in timings of phenomena, which are within the limitations of the physical models and the applied nodalizations. It was one objective of this investigation to evaluate the Severe Accident Management (SAM) procedures for VVER-1000 reactors, by for instance estimating the time available for taking appropriate decisions and preparing counter-measures. To evaluate the effect of possible operator actions, a SAM procedure (primary side depressurization) is included into the simulation. Without SAMs, the simulation provides plastic rupture of the RPV after approximately 4.3 h, while with SAMs, a prolongation of the vessel failure time is obtained by approximately 90 minutes. Currently, the late phase of the accident is investigated in more detail by comparing the lower head behaviour as simulated by ASTEC with results from dedicated finite element calculations. The work contributes to the reliability of the ASTEC code by means of plant applications.
An Approach to Controlling Inaccessible Regions for Inspection Defined by Rules on Fitness-For-Service in Japan
Authors: Ota Takashi , Dozaki Koji
Published: February 2010

On September 2007, Primary Water Stress Corrosion Cracking (PWSCC) flaws were found on the dissimilar metal weld of the steam generator (SG) inlet pipe nozzles of Tsuruga-2. Following the Tsuruga-2, similar cases were found in some other plants. These cracks were located in inaccessible regions by Ultrasonic Testing (UT) examination applied from the outer surface. Triggered by these cases, the Nuclear and Industrial Safety Agency (NISA), Japanese regulator of nuclear industries, directed nuclear power plant owners to investigate current status of these inaccessible regions for inspection defined by rules on fitness-for-service in Japan, and required to show developing plan on alternative measures of inspection for UT-exempted welds. On the other hand, the study to manage inaccessible regions in the rules on fitness-for-service has been started. The authors consider and propose a possible approach for modified rules of inspection in order to make control of these inaccessible regions.
Flow Regime Classification and Transition of Flow Boiling Through Porous Channel
Authors: Bai Bofeng , Zhang Xiaojie , Liu Maolong , Su Wang
Published: February 2010

In the present research, a visual experiment was carried out on the flow regimes of the porous channels in the spherical fuel element nuclear reactor. Boiling two-phase flow in different porous channels composed of particles with diameters of 4mm, 6mm and 8mm were studied respectively, and four different flow regimes occur within the experimental parameter range: bubbly flow, bubbly-slug flow, slug flow and slug-annular flow. The effects of heat flux, mass flux and particle diameter on the flow regimes were obtained. Bubbles and slugs deform, coalesce and break up more frequently, and increase in both number and size with the increase of the heat flux; bubbles and slugs tend to decrease in number and size with higher mass flux and particles of smaller diameters. At higher mass flux, a higher heat flux is needed to get the same flow regime that occurs at lower mass flux; with particles of smaller diameter, a higher heat flux is needed to get the same flow regime as that of particles of larger diameter. The flow regime map and flow regime transition have been proposed by modifying the void fractions of Tung/Dhir model.
The Velocity and Temperature Measurement of a Water Flow in the Wire-Wrapped 7-Rods Bundle System in FBR
Authors: Ochi Daisuke , Someya Satoshi , Okamoto Koji
Published: February 2010

Phosphorescence has been applied to measure the temperature. The intensity and the lifetime of phosphor depend on the temperature due to the quenching effect of the phosphor. In the present study, the temperature sensitive particles (TSParticles) with metal complex, phosphor molecules, were synthesized. TSParticles can be also particles for the particle image velocimetry. The particles were illuminated by a pulse laser at 20Hz. A high speed camera was applied to record 30 particle images at intervals of 25∼50 micro seconds (20∼40kHz) for each excitation laser pulse. From each series of images the velocity/temperature field was calculated. The TSParticle technique was applied to measure the velocity/temperature distribution of a water flow in a wire-wrapped rod bundle system, which briefly simulated the fuel rods system in FBR. The working fluid was water and the aim of this study was to contribute to an estimation and a certification of results of numerical simulation for the safety design of the FBR.