PIV Measurement of Pressure Distributions about Single Bubbles

Measurements of velocity and pressure distributions around a bubble are of fundamental importance to model the forces acting on the bubbles and to verify detailed numerical methods for the prediction of flow in nuclear reactors. The measurements of velocity distributions around a bubble have been conducted to understand the interaction between liquid flow and bubbles. However there are few studies on pressure distributions around a bubble for the lack of measurement method. In this study, we developed a method for evaluating a pressure distribution by making use of velocity data obtained by a particle image velocimetry (PIV) or a particle tracking velocimetry (PTV), and applied it to laminar pipe flows, laminar flows around single particles and single bubbles in a pipe to examine its accuracy and applicability to the flow around single bubbles. As a result, we could confirm that the method can evaluate the pressure distribution in various laminar flows, provided that the velocity data possess a good quality and a flow of concern is two-dimensional. The proposed method therefore has a potential to provide the important information for modeling of the bubble motion and verification of CFD methods such as interface tracking and lattice Boltzmann methods.     

Stilbene and NE 213 Characteristics Related to Their Use for Neutron Spectroscopy

The pulse discrimination technique permits scintillators to be used for neutron spectroscopy, in the presence of gamma rays, to investigate a wide variety of problems. The influence of shape and size of stilbene and NE 213 scintillators on their gamma-ray and neutron efficiencies and on the shape correction has been investigated. The neutron efficiency and the shape correction factor enter into the spectrum conversion function which is needed to reduce a recoil-proton energy distribution to a neutron spectrum. Gamma-ray and neutron efficiencies from 0.1 to 24 MeV have been graphed as a function of length. Graphs of the shape correction factor and the spectrum conversion function, up to 24-MeV neutron energy, have been made for various dimensions and similar volumes. These graphs and the summary tables can be used to select the size and shape of a detector to minimize the spread of the correction factor over a given range of neutron energy and to increase the neutron efficiency consistent with conditions of a given experiment or application.     

Measurement of the subpixel structure of AXAF CCD's

The authors present a method to measure the subpixel structure of a charge-coupled device (CCD), information necessary to accurately determine (<1% uncertainty) the absolute detection efficiency of the device. Their approach uses a thin metal film with periodically spaced holes (small, compared to the pixel size) to localize incident X-rays to a particular region of the pixel. The mesh is rotated to create a small angular misalignment between the grid holes and the CCD pixels, producing a moire effect in the data. The resultant moire pattern is compared to a CCD model, and a best fit minimization technique is used to constrain the parameters that describe the subpixel structure. This technique was developed to measure and calibrate the X-ray CCD's that will comprise one of the two focal plane instruments on-board AXAF, but it is applicable for measuring the structure of any pixelated solid state device     

A finite element formulation for thermal stress analysis. Part II: Finite element formulation

This paper deals with the development of a variational principle which can be used for solving problems related to the thermoelastic behavior of solids and is the first of the two part series. The formulation is based on the introduction of a new quantity defined as heat displacement and related to temperature in the same manner as the mechanical displacement is related to strain. The introduction of such a quantity allows the heat conduction equations to be written in a form equivalent to the equation of motion, and the equations of coupled thermoelasticity to be written in a unified form. The obtained equations are used to write a variational formulation which, together with the concept of generalized coordinates, yield a set of differential equations with the time as the independent variable. These equations can be used to formulate a finite element solution for thermoelastic problems. This is done in the second part.     

Experimental Investigation of the Steam Condensation Phenomena due to a Multi-hole Sparger

KAERI has performed a series of steam condensation tests to assess the performance of a unit cell sparger that will be used in the APR1400 reactor. A simplified I-sparger was used for the steady state steam condensation tests to study the characteristics of the condensation phenomena due to a multi-hole sparger and to provide test data for a code development and verification. A range of steam mass fluxes for the steady state condensation tests were selected to define the transition region from the condensation oscillation regime to the stable condensation regime. Condensation loads and a variation of the frequencies of the pressure waves due to a steam condensation are analyzed. In addition, the local temperature distribution near the sparger discharge holes is discussed and a condensation regime map for a multi-hole sparger has been suggested.     

基于中子平衡研究增殖燃料实现CANDLE模式的最优配置

寿期内中子通量、核素浓度和功率分布的轴向形状均保持恒定(Constant Axial shape of Neutron flux,nuclide densities and power shape During Life of Energy produced,CANDLE)是实现原位增殖-焚烧(Breed-and-Burn,BB)模式的一种燃耗策略。CANDLE堆经易裂变燃料或外中子源进行点火,启动后由增殖燃料的燃烧实现自稳运行。若要CANDLE堆自稳运行于k_(eff)=1,必须对堆芯几何及燃料体积分数进行配置优化。最优配置方案可通过蒙特卡罗方法模拟CANDLE堆芯,根据有效增殖因子筛选得出。但该方法需耗费大量的计算时间,若采用1D模型近似模拟,并结合中子平衡方法进行分析,便可大幅节约计算时间,获得具有指导性意义的结果。本文将论证该方法的可行性,并应用该方法估算钠冷贫铀CANDLE堆半径在100 400 cm、燃料体积分数在35%60%变化时的最优配置。     

Development of a risk-informed accident diagnosis and prognosis system to support severe accident management

RISARD, risk-informed severe accident risk diagnosis system, is a computerized tool developed to improve a severe accident management (SAM) for a nuclear power plant and to effectively support the MCR and the TSC in executing the relevant SAM activities. In order to provide a diagnostic capability to a state of the plant and a prognostic capability for an anticipated accident progression, the system examines (a) a symptom-based diagnosis of a plant damage state (PDS) sequence in a risk-informing way and (b) a PDS sequence-based prognosis of key plant parameter behavior, through a prepared database (DB) containing plant-specific severe accident risk (SAR)-related information. For a given accident, the replicated use of these two processes makes it possible to obtain information about the functional states of the plant and containment safety systems expected at the time of a severe accident as well as future trend of the key plant parameters that are essentially required for taking the relevant SAM action, eventually leading to an answer about the best strategy for SAM. The foregoing concept for an accident diagnosis and prognosis can give the SAM practitioners more time to take action for mitigating the consequences of the potential accident scenarios since they are made in a simple, fast, and efficient way through a prepared SAR database and it is useful especially when the plant information available for SAM is incomplete and limited. The main purpose of this paper is to (a) introduce the concept of the RISARD system proposed to support SAM and its implementation through a prepared OPR1000 plant- and MAAP code-specific SAR database and (b) assess prediction capabilities of major events expected during the evolution of a severe accident through the system.     

Real time application of simultaneous estimation theory for parameters and state to one-dimensional heat conduction

This paper applies filtering theory of the parameters and states of a distributed system to one-dimensional heat conduction in order to obtain information on the real time application of nonlinear estimation theory. This involves a trial to study a simple and fast algorithm for the state estimation of power reactors limited to a small number of sensors by noisy observations. The authors estimated the heat transfer coefficients and the transient temperature profile in a copper rod simultaneously in real time by using a small computer, NOVA 3/12, where the parameter and state of the heat conduction-type P.D.E. could serve as a simplified alternative to the nuclear constant and power profile. The authors also examined the feasibility of a nonlinear filter provided for simultaneous estimation, considering unknown parameters as alternative state variables. The filter tested converged during the sampling period and generated reasonable estimates of the parameter and state from noisy observations with a small number of sensors and at a discrete number of times.     

Finding an optimization of the plate element of Egyptian research reactor using genetic algorithm

The second Egyptian research reactor ET-RR-2 went critical on the 27th of November 1997. The National Center of Nuclear Safety and Radiation Control （NCNSRC） has the responsibility of the evaluation and assessment of the safety of this reactor. The purpose of this paper is to present an approach to optimization of the fuel element plate. For an efficient search through the solution space we use a multi objective genetic algorithm which allows us to identify a set of Pareto optimal solutions providing the decision maker with the complete spectrum of optimal solutions with respect to the various targets. The aim of this paper is to propose a new approach for optimizing the fuel element plate in the reactor. The fuel element plate is designed with a view to improve reliability and lifetime and it is one of the most important elements during the shut down. In this present paper, we present a conceptual design approach for fuel element plate, in conjunction with a genetic algorithm to obtain a fuel plate that maximizes a fitness value to optimize the safety design of the fuel plate.     

The UWMAK-II study: A conceptual design of a helium cooled, solid breeder, Tokamak fusion reactor system

UWMAK-II is a conceptual design study of a low ß, circular Tokamak fusion power reactor. The aim of the study has been to perform a self-consistent analysis of a probable future fusion power system based on the philosophy that design decisions, wherever possible, should be conservative and should be based on present technology. As such, this system will not be the smallest, the least expensive, or the optimum Tokamak reactor. Rather, it represents a feasible system which we use to assess the technological problems uncovered and to examine possible solutions. The plasma is designed to generate 5000 MW(th) during a pulse and 1709 MW(e) continuously based upon a burn cycle with a 90 min burn and a 6.5 min rejuvenation period. The plasma carries a current of 14.9 MA and is designed with a double null poloidal divertor for impurity control and particle pumping. In addition, a low Z liner in the form of a carbon curtain is included to eliminate any source of high Z impurities. Plasma heating to ignition involves the use of neutral beam heating for a 10 sec period during which 200 MW of 500 keV deuterium atoms are injected into the plasma.The blanket design employs helium cooling and the solid lithium-bearing compound, lithium aluminate (Li₂Al₂O₄) for breeding tritium. The structural material is 316 stainless steel and beryllium is used as a neutron multiplier. The neutron radiation environment produces radiation damage that considerably influences blanket and system performance. The most significant effect is the loss of ductility which appears to limit the usable lifetime of the blanket first wall to about 2 yr at a 14 MeV neutron wall loading of 1.16 MW/m². The solid breeder blanket minimizes the tritium inventory but because of the low fractional burnup in the plasma and the need for roughly a one day reserve of fuel, the inventory is 17.7 kg. Induced radioactivity levels in the structure are of the order of 1 Ci/W(th) at shutdown after two years of operation. The main contributors to the activity are ) and ). Afterheat levels are slightly above 1% of thermal power but the afterheat power density is low, less than 0.1 w/g. The power cycle involves a He---Na intermediate heat exchanger followed by a sodium—steam system. The sodium intermediary is used to minimize tritium leakage through the power cycle and to provide a working fluid for thermal energy storage such that continuous electrical output is produced despite a pulse plasma cycle. A materials resource study has been completed for a UWMAK-II type system and beryllium appears to present a particular problem with regard to adequate resources. Other materials that could present problems of procurement include chromium and nickel. A preliminary economic analysis has been carried out to identify major cost areas and this is described.     

Copper canister lifetime limited by a sulfide intrusion in a deep geologic repository

Korea has continuously implemented an ambitious nuclear energy deployment program since 1978. Korea currently operates 20 units, 16 PWRs and four CANDUs and constructs four and reviews license application of two more units. Also, Korea plans to build two more units by 2016. In addition, according to the new “Green Growth Plan while reducing the emission of carbon dioxide” Korea will introduce 10–12 units by 2030. This will inevitably result in more burdens on the safe management of spent nuclear fuels. Korea Atomic Energy Research Institute has developed a final disposal concept for Spent Nuclear Fuel (SNF) named KRS. KRS proposes to emplace SNF in a deep geologic formation such as a crystalline rock. Two key engineered barriers are applied to retard the potential release of a radionuclide from an embedded SNF; a waste container and an engineered barrier. Such an engineered barrier is composed of domestic calcium bentonite and the waste container is composed of an outer copper layer and an inner steel layer. The outer layer, a copper layer is dedicated to protect a waste container against corrosion. The main corrosion mechanism to corrode a copper waste container is a pitting whose speed of corrosion is 5–25 times higher than that of a uniform corrosion. In this paper, a special mass transfer resistance model is developed to predict the migration of sulfide from a fracture to a waste container surface via a bentonite layer. Based on it the lifetime of a copper canister layer limited by a pitting corrosion is estimated. Results show that under normal conditions, a copper layer can sustain its integrity for up to more than millions of years.     

Advanced DVI for ECC direct bypass mitigation

An ECC direct bypass fraction during a late reflood phase of a LBLOCA is strongly dependent on the characteristics of the cross flow and the geometrical configuration of a DVI in the downcomer of a pressurized light water reactor. The important design parameters of a DVI are the elevation, the azimuthal angle, and the separator to prevent a steam-water interaction. An ECC sub-channel to separate or to isolate an ECC water from a high-speed cross flow is one of the important design features to mitigate the ECC bypass phenomena. A dual core barrel cylinder as an ECC flow separator is located between a reactor vessel and a core barrel outer wall in the downcomer annulus. A new narrow gap between the core barrel and the additional dual core barrel plays the role of a downward ECC flow channel or an ECC flow separator in a high-speed cross flow field of the downcomer annulus. The flow zone around a broken cold leg in the downcomer annulus has the role of a high ECC direct bypass due to a strong suction force while the wake zone of a hot leg has the role of an ECC penetration. Thus, the relative azimuthal angle of the DVI nozzle from the broken cold leg is an important design parameter. A large azimuthal angle from a cold leg to a hot leg needs to avoid a high suction flow zone when an ECC water is being injected. The other enhancing mechanism of an ECC penetration is a grooved core barrel which has small rectangular-shaped grooves vertically arranged on the core barrel wall of the reactor vessel downcomer annulus. These grooves have the role for a generation of a vortex induced by a high-speed cross flow. Since the stagnant flow in a lateral direction and rotational vortex provides the pulling force of an ECC drop or film to flow down into the lower downcomer annulus by gravity, the ECC direct bypass fraction is reduced when compared to the current design of a smoothed wall. An open channel of grooves generates a stagnant vortex, while a closed channel of grooves creates an isolated ECC downward flow channel from a high-speed lateral flow. In this study, new design concepts for a dual core barrel cylinder, grooved core barrel, and a reallocation of the DVI azimuthal angle are proposed and tested by using an air-water 1/5 scaled air-water test facility. The ECC direct bypass reduction performances of the new design concepts have been compared with that of the standard type of a DVI injection. The azimuthal angle of the DVI nozzle from a broken cold leg varies from −15° to +52° toward a hot leg. The test results show that the azimuthal injection angle is an effective parameter to reduce the ECC direct bypass fraction. The elevation of the DVI nozzle is also an important parameter to reduce the ECC direct bypass fraction. The most effective design for reducing the ECC direct bypass fraction is a dual core barrel. The reduction fraction when compared to the standard DVI is about −30% for the dual core barrel while it is −15% for the grooved core barrel.     

Deployment analysis and remote accessibility verification for a maintenance task in a PRIDE digital mock-up

This study introduces an advanced modeling and simulation system that can verify and optimize a maintenance procedure at the early design stage of a virtual engineering system. System architecture, which is composed of modules used to analyze deployment of devices that deal with radioactive materials in a digital mock-up, and modules used to simulate accessibility of a remote manipulator for maintenance tasks, are discussed. Based on this architecture, new technology that enables a maintenance analyst to analyze pyroprocessing is proposed. Workspace analysis for remote manipulators, which can optimize a maintenance task at the radiation control area, is introduced. The mathematical background, with respect to the forward and inverse kinematics for a virtual manipulator accessing devices in a virtual environment, is described to establish a remote accessibility. Virtual prototyping is illustrated to carry out an experiment on the deployment analysis of devices and remote accessibility by a haptic device.     

Shape optimization of a perforated pressure vessel cover under linearized stress constraints

One of the general methods to evaluate a failure condition is to compare a maximum stress with an allowable stress. A failure condition for a stress is usually applied to a concerned point rather than a concerned section. In an optimization procedure, these stress conditions are applied as constraints. But the ASME code that prescribes its general rules upon the design of a NSSS (nuclear steam supply system) has quite a different view on a failure condition. According to the ASME code Sec. III, a stress linearization should be performed to evaluate a failure condition of a structure. Since a few programs provide a procedure for a stress linearization through a post-processing stage, an extra calculation of the linearized stresses and the derivatives of a linearized stress are conducted to adopt the stress linearization results to an optimization procedure as constraints. In this research, an optimization technique that utilizes the results of a stress linearization as a constraint is proposed. The proposed method was applied to the shape design of a perforated pressure vessel cover.     

The Application of Visualization and Simulation in a Dismantling Process

The lack of as-built drawings and information for many old nuclear facilities impedes the planning for a decommissioning. Traditional manual walkthroughs subject workers to lengthy exposure to radiological and other hazards. And also there have been a delay in the dismantling schedules and that in an increase in the amount of waste, which have caused a heavy expenditure for the decommissioning cost. In order for a 3D model to represent the real dismantling operation environment, realistic dismantling activities and scenarios were modeled. We present a dismantling digital mock-up system that can show a dismantling process through an animation and enable us to evaluate the major parameters that can directly affect the cost in a static simulation as a preview to the dismantling activities on a hypothetical dismantling environment. We carried out a prototype 3D animation and a simulation system for dismantling the thermal column in Korean Research Reactor-1(KRR-1). The results show that the modeling of a realistic scenario makes it possible to demonstrate the adequacy of the design and can provide operators with a high fidelity under the real working conditions.     

Analysis of Flows around a BWR Spacer by the Two-Fluid Particle Interaction Method

The Moving Particle Semi-implicit (MPS) method, a particle interaction method developed in recent years, is formulated by representing the differential operators in Navier-Stokes equations as the interaction between particles characterized with a kernel function and adopts a mesh-free algorithm. The MPS method is particularly suitable for treating liquid breakup. We extended the MPS method to a two-fluid system, introduced a potential-type surface tension, and developed a kernel function for the interface between liquid and gas to simulate two-phase flows. This extended MPS method, which we call Two-Fluid MPS (TF-MPS) method, has been verified through a number of analyses of two-phase flow experiments. The objectives of this study are to verify the applicability of TF-MPS method to a flow around a BWR spacer and to make up constitutive correlations for macroscopic methods. In this paper, we describe the formulation and the calculation algorithm of TF-MPS method, and present the results of the verification studies.     

Pressure release of containments during severe accidents in Switzerland

As required by the Swiss Federal Nuclear Safety Inspectorate (HSK) all Switzerland's five nuclear power plants have to install a containment filtered venting system. The integrity of the containment (the last barrier for radioactive releases to the environment) can be threatened by overpressure due to inadequate heat removal. Design requirements have been provided for a specific class of severe accident scenarios. In general the capacity of the system is considered sufficient if it is able to vent the steam production corresponding to a decay heat level of 1% of the thermal reactor power. The mitigation capacity for the reduction of released radioactive material is specified by a retention factor of 1000 for aerosols to prevent or limit a long term ground contamination and a factor of 100 for elementary iodine for prevention or limiting of thyroid doses and to avoid short term evacuation. Besides existing requirements for design, maintenance and operation, additional claims such as passivity and operability at any pressure conditions inside the containment have to be met. Passivity implies that the system can be initiated after a severe accident without any operator action. The system also has to allow early manual venting. Various filtered venting systems are presently available. The nuclear power plants of Beznau, Gosgen, Leibstadt and Muhleberg have already selected such systems and already implemented them or are going to install them step by step. Beznau selected the Sulzer-EWI system which is using a water pool with nozzles-baffle plates and mixing elements to achieve the required filtration of the aerosols. In both Beznau units, the systems are installed and in standby mode. Gosgen, a pressurized water reactor as well as Beznau, is going to implement a filter system developed by Siemens-KWU, known as sliding pressure venting process, combining a venturi scrubber in a water pool and a mesh filter. The boiling water reactor of Leibstadt also selected the same system as Beznau while Müheberg choose the ABB system but not in the common design. The venturi pipes are thereby integrated in the water pool of the outer torus. The system in all five nuclear power plants is fully operable and in standby mode since December 1993.     

An evaluation of nuclide release from a trench-type LILW repository

A simple and effective model and a GoldSim (GoldSim, 2006) template program, by which a probabilistic safety assessment of a conceptual trench-type repository for low- and intermediate level radioactive waste (LILW) disposal can be carried out under various nuclide release scenarios, have been developed. To quantify the exposure dose rates due to nuclide release from the trench and transport through the various pathways possible in the near- and far-fields of the repository system under a base case and some alternative scenarios, illustrative evaluations for a comparison among the scenarios as well as a sensitivity of shortcut pathways generated due to earthquakes on the nuclide transport are made and demonstrated. To this end, by changing the conservative base case nuclide release scenario under which all portions of the cap of the trench are failed unconditionally and immediately after a closure of the repository, a total of four other alternative scenarios were separately evaluated for the total exposure dose rates to the farming exposure group and then compared to the base case results. Among them, an earthquake scenario shows a dominant behavior almost throughout the whole time span. To see the influence of all shortcut travel pathways that are assumed to be newly generated by this earthquake scenario, their sensitivities to the exposure dose rates to the farming exposure group were also made and compared to each other.     

Study on Solid-Liquid Two-Phase Flow on PWR Sump Clogging Issue

It has been a concern that sump screen clogging would occur in pressurized water reactors (PWRs) in the case of a loss-of-coolant accident (LOCA), because two-phase jet flow would strip off thermal insulation from the piping and wash down the broken and fragmented debris to sump screens. It is necessary for the evaluation of the effectiveness of sump screens to estimate the amount of transported debris from a break position to sumps. In general, conservative logic trees have been used to determine debris transport rates. Realistic debris transport evaluation is useful for considering measures and rational decision making in licensing. The purpose of this study is to develop a debris transport evaluation model and to apply the model to this issue. We developed a solid-liquid multiphase model that is capable of simulating debris transport, settling, and resuspension. The model is able to treat solid particles of different sizes, which are smaller than uniform-sized liquid particles. This approach contributes to reducing the calculation cost in a large-scale simulation. The model and a turbulence model were implemented into a code based on the moving particle semi-implicit (MPS) method. Several open-channel hydraulic experiments with fibrous debris were conducted. The code named SANSUI 2.0 was validated by the comparison of the analytical results with experiments. This method was applied to the debris transport analysis of a full-scale PWR containment vessel floor, and the debris transport behavior was evaluated.