bluuebunny/crossref_metadata_2025 huggingface dataset:
The Local Governments’ Management of the Sensitive Information
Authors: Sugawara Shin-etsu , Inamura Tomoaki , Madarame Haruki
Published: April 2011

After the September 11th terrorist attacks, the Convention on the Physical Protection of Nuclear Material was revised, and many countries have enhanced their regulatory regimes about the management of sensitive information, especially in the physical protection system. Japan also amended the Nuclear Reactor Regulation Law in 2005 in step with this global movement. The major areas of this revision which are associated with sensitive information are as follows: formulation of the Design Basis Threat (DBT), introduction of inspection system of physical protection and obligation of confidentiality of the secret of physical protection. Through this amendment, the responsibilities of the national government and the utilities have been clarified. However, there is no prescription which ordains the role and responsibility of the local governments. In fact, the local governments receive various information from the utilities through the “Safety Agreements” which are concluded between the local governments and the utilities, and the Public Safety Commissions of prefectures are involved in the transportation of nuclear materials. Moreover, the Act on Special Measures concerning Nuclear Emergency Preparedness provides the engagement and the responsibility of the local governments in case of nuclear disaster. In addition, the Civil Protection Law also provides the formulation of local governments’ plans for a response to national emergencies including nuclear disaster which is caused by terrorist attacks. As described above, the local governments are in a position where they can or have to touch the sensitive information in a variety of ways. Originally, the local government employees have obligation of confidentiality by the Local Public Service Act. Thus, about the sensitive information, they have duty to keep secret. However, we are hard to say that there are complete systems to check this obligation, so we can point out that its effectiveness is doubtful. Especially, the sensitive information which is related to nuclear materials is vital for security of the nation as a whole. Under such awareness, we’re studying the change of the local governments’ way of the management of sensitive information accompanied by the strengthening of Japanese nuclear regulation, and the actual condition of it. Now, we interview some local governments’ departments in charge where nuclear facilities are located. In this paper, we discuss the actual condition and the problems around the local governments’ management of the sensitive information.
Numerical Research on 3-D Flow Field in 5×5 Rod Bundles With Spacer Grids
Authors: Tian Rui-feng , Mao Xiao-hui , Gao Pu-zhen , Wang Xiao-jun
Published: April 2011

3-D single-phase flow field in 5×5 rod bundles with spacer grids was studied by numerical method. Using hybrid grids technique, SST k-ω model and SIMPLEC algorithm, the Reynolds averaged mass conservation and momentum conservation equations were solved, and the pressure and velocity field were obtained. The simulation results show that the spacer grids leads to intense lateral flow in rod bundles channel, and it submitted parabola distributing raw; axial velocities were distributed uniformly in the channel; drag coefficient decreased as inlet Reynolds number increased. Calculated results agreed with experiment data well, it shows that numerical methods employed in this paper is suitable to study flow field in 5×5 rod bundles.
Study and Design of Monitoring System of Reactor Coolant Pump in Nuclear Power Plant
Authors: You Lei , Sun Fuchun , He Pan , Xu Hongkun , Fang Fang
Published: April 2011

In this paper, we develop a monitoring system of reactor coolant pumps in nuclear power plant (CPS). The safe running of reactor coolant pump is important for nuclear power plant. Based on the Fourier transform (FT) and some algorithm, The data collected from the pump are analyzed. Once the accident happens, it would cause unimaginable outcome. The system will be jumped to failure process mode when the pump has something wrong. The advanced VXI and virtual instrument technology are applied to system, and the reactor coolant pump will be monitored overall so as to assure that the reactor coolant pump runs in safe, which has a significant value to secure the safe operation and reliability of the nuclear plant. The monitoring system will help the operators find fault of reactor coolant pump.
Study of Weld Residual Stress Field in the Girth Seam H6A of Core Shroud of Boiling Water Reactor
Authors: Li Yongkui , Kaji Yoshiyuki , Igarashi Takahiro
Published: April 2011

Many accidents have occurred in nuclear power plants due to the intergranular stress corrosion cracking (IGSCC) in the heat affected zone (HAZ) of welded joint of the core shroud of boiling water reactors (BWRs) in past years. The IGSCC is considered to be caused by the synergistic roles of corrosion environment, neutron irradiation and the welding residual stress. After several decades, the degradation of Type 316L low carbon stainless steel used in the core shroud occurs due to the neutron irradiation and thermal cycles. The degradation can be referred to the irradiation hardening, segregation of the local chemical composition at grain boundaries and swelling. The synergistic effects of those eventually lead to the initiation and propagation of the irradiation-assisted stress corrosion cracking (IASCC) in core shroud for long operation. The HAZ of the girth seams H6a in the core shroud are sensitive to the stress corrosion cracking. We are focusing on the weld residual stress field around the girth seam H6a in the core shroud as weld. The analysis work adopted different approaches in ABAQUS to simulate the weld residual stress, and they are Static General Analysis (SGA) and Fully Coupled Temperature-Displacement Analysis (FCTDA) respectively. The former is much simple to finish the progress while cannot obtain much accurate results at the boundaries of beads due to the discontinuous temperature field in the model. The later analysis gave the much accurate results comparing with the experimental results. The axial stress field in the crossing section of the wall of the core shroud was also clarified.
Frame of Two Way Risk Communication by Learning Organization to Foster Safety Culture
Authors: Yoshikawa Hidekazu
Published: April 2011

Public participation to the decision of nuclear issues has been progressing in the already developed countries of nuclear power such as USA, West Europe and Japan. Risk communication by the nuclear industries to the society is time consuming and high cost. The authors of this paper have proposed the frame of two way risk communication for the nuclear industries to cope with the difficulty of social acceptance. In this paper, the author investigated the case of the largest nuclear power station in Japan which had hit by large earthquake in July 16, 2007, being back to the normal operation, with continuous plant restoration work and hard efforts to get approval from local government of commercial operation, from this frame of two way risk communication.
Corrosion Test of 304 Stainless Steel in Hydrazine and Methanol Solutions at 320°C Under Gamma-Irradiation and Gamma Radiolysis of Hydrazine and Methanol
Authors: Domae Masafumi , Kawamura Hirotaka , Akutagawa Daisuke , Hisamune Kenji , Katsumura Yosuke , Muroya Yusa
Published: April 2011

It has been pointed out that dissolved hydrogen is one of the key factors of PWSCC (primary water stress corrosion cracking) in the primary systems of pressurized water reactors. The authors consider that application of alternative reductant for hydrogen may mitigate PWSCC. The corrosion tests of 304 stainless steels in 2 mmol dm−3 methanol solution and 2 mmol dm−3 hydrazine solution at 320 °C were carried out under γ-ray irradiation, and the corrosion environment was evaluated. Electrochemical corrosion potential (ECP) of the stainless steels was measured during the immersion tests. The ECP values were −605 mV and −643 mV vs. SHE in 2 mmol dm−3 hydrazine solution and 2 mmol dm−3 methanol solution at 320 °C, respectively. After the immersion tests, oxide films formed on the stainless steel specimens were analyzed with scanning probe microscope and X-ray diffraction. It is concluded that from the ECP measurement and comparison with previous results corrosion environment under following conditions is similar: (1) DH 1.5 ppm without irradiation, (2) methanol 2.9 ppm without irradiation, (3) hydrazine 2 mmol dm−3 under γ-ray irradiation and (4) methanol 2 mmol dm−3 under γ-ray irradiation. During the immersion tests, the test waters were sampled several times, and analyzed, in order to understand radiolysis of the methanol solution and the hydrazine solution at 320 °C. Hydrazine is decomposed predominantly through thermal decomposition, and an ammonia molecule is formed from a hydrazine molecule. Formaldehyde and ethylene glycol were detected in the methanol solution. But, carboxylates were not detected. The present results do not demonstrate oxidation of methanol to CO or CO2. It is pointed out that irradiation experiments of higher absorbed dose are necessary.
Dialogue Process Between Expert and Public Regarding HLW Disposal in Japan
Authors: Furukawa Masashi , Kimura Hiroshi , Madarame Haruki , Sugiyama Daisuke , Chida Taiji
Published: April 2011

Different people have different opinion on nuclear power use. As for HLW disposal, more complex issues such as uncertainty due to very long time-scale required for disposal would add to the difficulty of conducting smooth communication and building stable consensus between the scientific experts and the general public. We carried out Q&A dialogue experiments between an expert and public participants who were indifferent to HLW disposal. The dialogue process was divided into two periods. In the learning period, participants learned fundamental knowledge about HLW disposal. In the discussion period, participants discussed about one specific topic, such as “Resource and Energy”, “Geological Disposal”, or Safety Violation. These dialogue experiments can help the experts to communicate and conduct comprehensive activities with the public which would help implement the HLW disposal from the public point of view.
Coefficients for I-131 Sorption on Different Surfaces
Authors: Stempniewicz M. M.
Published: April 2011

Sorption of fission product vapors on metallic surfaces and dust particles is an important safety aspect of HTR reactors. Safety analyses of these reactors are performed using computer codes, such as MELCOR, RADAX, SPECTRA. These codes have sorption models allowing to compute the sorption rates of different fission products on surfaces. The code users must supply the model coefficients applicable for the particular surface and isotope. This paper describes the work performed to find relevant experimental data and find the sorption coefficients that represent well the available data for iodine on different surfaces. The purpose of this work is to generate a set of coefficients that may be recommended for the computer code users. Calculations were performed using the computer code SPECTRA. The following data was analyzed: • Sorption of I-131 on graphite; • Sorption of I-131 on steel; • Sorption of I-131 on dust. The results are summarized as follows: • The available data is provided in form of Langmuir isotherms. • The Langmuir isotherms do not provide sufficient data to define all sorption coefficients. The Langmuir isotherm provides equilibrium data; the relaxation time (to get to equilibrium) needs to be guessed. In practice this means that one of the sorption coefficients must be guessed. In the present calculations the desorption coefficient was being guessed and then varied in sensitivity calculations. The calculations showed that surface concentration is not sensitive to the choice of the parameter. • The sorption model in SPECTRA is capable to correctly reproduce the sorption behavior given by the Langmuir isotherms. • Out of the calculated cases, the highest activity (surface concentration) is observed on the steel surface; the lowest on the graphite surfaces. • The present work may serve as a useful guide of how to convert the Langmuir isotherm data into the input parameters required for computer code calculations.
Analysis of Steady State Temperature Field of Rod-Shaped Fuel Element Based on Virtual Reality
Authors: Zhu Bo , Yang YanHua , Chen Wenzheng , Li Shiting
Published: April 2011

Based on the reactor physics and heat transfer parameter of rod-shaped fuel element, the steady state distribution and variation for rod-shaped assembly are analyzed and computed; using the virtual reality technology, 3D geometric entities of fuel element is constructed and divided into grid according to different condition; Through the normalized treatment for numerical results, the VR simulation environment for temperature field of fuel element is constructed. The research for field visualization has some reference value for core thermal design and operational safety analysis.
Cogeneration of Clean Water Using Nuclear Power Plant Waste Heat
Authors: Chen Kuan , Sandquist Gary M. , Chun Wongee
Published: April 2011

The production of clean water in the US as well as other countries is a critical need along with non-greenhouse gas electrical power generation. Low-temperature waste heat from nuclear power plants can be used to produce the large quantities of clean water for reactor cooling (∼25,000 acre-ft/yr), potable water for culinary and agricultural use and many other applications. Cogeneration of nuclear electrical power and clean water is reviewed and discussed in this paper. These cogeneration systems can utilize grey and/or brackish water that can markedly extend potential sites for future nuclear plants in areas where only poor water sources are available. A steam adsorption system for on-line production of clean water and refrigeration using nuclear power plant waste heat is also proposed and discussed. This improved design for more energy-efficient use of the steam adsorption cooling has the potential to substantially reduce the intense electrical power consumption for food processing and storage, ice- and snow-making and air-conditioning.
Subcooled Water Flow Boiling Heat Transfer in a Short SUS304-Tube With Twisted-Tape Insert
Authors: Hata Koichi , Masuzaki Suguru
Published: April 2011

The subcooled boiling heat transfer (HT) and the steady-state critical heat fluxes (CHFs) in a short SUS304-tube with twisted-tape insert are systematically measured for mass velocities (G = 4016 to 13850 kg/m2s), inlet liquid temperatures (Tin = 285.82 to 363.96 K), outlet pressures (Pout = 764.76 to 889.02 kPa) and exponentially increasing heat input (Q = Q0exp(t/τ), τ = 8.5 s) by the experimental water loop comprised of a multistage canned-type circulation pump controlled by an inverter. The SUS304 test tube of inner diameter (d = 6 mm), heated length (L = 59.5 mm), effective length (Leff = 49.1 mm), L/d (= 9.92), Leff/d (= 8.18) and wall thickness (δ = 0.5 mm) with average surface roughness (Ra = 3.18 μm) is used in this work. The SUS304 twisted tape with twist ratio, y [= H/d = (pitch of 180° rotation)/d], of 3.39 is used. The relation between inner surface temperature and heat flux for the SUS304-tube with the twisted-tape insert are clarified from non-boiling to CHF. The subcooled boiling heat transfer for SUS304-tube with the twisted-tape insert is compared with our empty SUS304-tube data and the values calculated by our and other workers’ correlations for the subcooled boiling heat transfer. The influences of the twisted-tape insert and the swirl velocity on the subcooled boiling heat transfer and the CHFs are investigated into details and the widely and precisely predictable correlations of the subcooled boiling heat transfer and the CHFs for turbulent flow of water in the SUS304-tube with twisted-tape insert are given based on the experimental data. The correlations can describe the subcooled boiling heat transfer coefficients and the CHFs obtained in this work within −25 to +15% difference.
Analysis of the EAST Cryogenic System Operational Modes
Authors: Hu Liangbing , Zhou Zhiwei , Yuan Chunyan , Zhuang Ming
Published: April 2011

For EAST operation schedule, auxiliary subsystems like the cryogenic system have to cope with different heat loads which depend on the different EAST operating states. The cryogenic system consists of a cryoplant and a cryodistribution system. All of these cryogenic subsystems have to operate in parallel to remove the heat loads from the magnet, 80K shields, built-in cryopumps and other small users. After a brief recall of the main particularities of a cryogenic system operating in a Tokamak environment, the first part of this study is dedicated to the assessment of the main EAST operation states. A new design of refrigeration loop for the HTS current leads, the updated layout of the cryodistribution system and revised strategy for operations of the built-in cryopumps have been taken into consideration. The relevant normal operating scenarios of the cryoplant are checked for the typical EAST operating states like plasma operation state, short term stand by, short term maintenance, or test and conditioning state. The second part of the paper is dedicated to the abnormal operating modes coming from the magnets and from those generated by the cryoplant itself. Thanks to this analysis, the optimization of the present operational modes is proposed to make match the technical specifications of the cryogenic system with the EAST operation requirements.
Simulation of Natural Circulation in Atmosphere of Experimental Containment Using Lumped-Parameter Model Based on CFD Calculations
Authors: Kljenak Ivo , Mavko Borut , Ursˇicˇ Mitja
Published: April 2011

The possibility to adequately simulate the flow circulation in the atmosphere of a nuclear power plant containment at accident conditions using a lumped-parameter code is investigated. An experiment on containment atmosphere mixing and stratification, which was performed in the TOSQAN experimental facility at IRSN in Saclay (France), was considered. During some phases of the experiment, steady states were achieved by keeping the boundary conditions constant. Two steady states during which natural convection was the dominant gas flow mechanism were simulated independently with the CFD code CFX4, whereas the entire transient was simulated with the lumped-parameter code ASTEC. The nodalisation of the lumped-parameter model was based on the flow pattern during the considered steady states, simulated with the CFD code. The lumped-parameter simulation succeeded in replicating the basic pattern of natural circulation in the vessel atmosphere during the two considered steady states, determined with the CFD code.
Heterogeneous Inherent Boron Dilution Transient: A CFD Analysis of a Hot Boron Depleted Slug Interacting With Colder Borated Water in a PWR
Authors: Freydier Philippe , Gaudron Bruno , Cornille Se´bastien , Lombard Virginie , Bertrand Pauline
Published: April 2011

For the study of the Heterogeneous Inherent Boron Dilution transient in a Pressurized Water Reactor, a Small Break Loss Of Coolant Accident (SB-LOCA) is postulated. Natural Circulation (NC) may be interrupted and, under Reflux-Condenser (RC) conditions, the steam formed in the core condensates in the Steam Generator (SG) U-tubes: a boron-depleted slug may accumulate in the crossover leg and in the SG outlet chamber. If NC restarts as the Reactor Cooling System (RCS) is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. The mixing of the boron depleted slug with the borated water in the Cold Legs (CLs), downcomer and lower plenum after Restart of Natural Circulation (RNC) is quantified by means of Computational Fluid Dynamics (CFD) analyses. The CFD code STAR-CD is used to perform this analysis. Boundary conditions for this calculation — especially the boron-depleted slug size and the NC restart mass flow rate — are extrapolated from PKL experimental findings. The initial conditions are derived from an overall plant analysis performed with the CATHARE system code. Buoyancy effects, both in the cold leg and in the downcomer, are very significant phenomena for the evaluation of the slug transport and mixing: the hot (saturation temperature) boron-depleted water slug tends to accumulate in the upper parts of the cold legs and in the upper part of the downcomer (above the cold legs), before being pushed and dragged down. The boron concentration distribution at the core inlet during the transient, evaluated with STAR-CD, is compared with a critical value in order to check that boron concentration at the core inlet is always above the threshold necessary for the core to remain subcritical.
The Development of a Generic Qualified Digital Platform
Authors: Shyu Victor S. S. , Chen Ming-Huei
Published: April 2011

The nuclear industry and research institutes in Taiwan are conducting a joint effort project to establish a self-reliant nuclear Instrumentation and Control (I&C) system design and fabrication capabilities in Taiwan. The purposes of this project, as called Taiwan’s Nuclear I&C System (TaiNICS), are planned to support digital upgrade of the existing nuclear power plants and the new nuclear installations in Taiwan. The project will be a long term pursuit of several task branches, including establishment of a generic qualified digital platform, qualification and certification processes, nuclear I&C systems design, safety analyses for software common cause failure, licensing, and collaboration. The short term goal of this project is to submit the License Topical Report (LTR) of a generic digital platform for the review of Taiwan’s regulatory body in 2013.
Acceleration Technique Using Krylov Subspace Methods for 2D Arbitrary Geometry Characteristics Solver
Authors: Zhang Hongbo , Wu Hongchun , Cao Liangzhi
Published: April 2011

The Generalized Minimal RESidual (GMRES) method, which is a widely-used version of Krylov subspace methods for solving large sparse non-symmetric linear systems, is adopted to accelerate the 2D arbitrary geometry characteristics solver AutoMOC. In this technique, a formulism of linear algebraic equation system for angular flux moments and boundary fluxes is derived as an alternative to traditional characteristics sweep (i.e. inner iteration) formalism, and then the GMRES method is implemented as an efficient linear system solver. Several numerical results demonstrate that the acceleration technique based on Krylov subspace methods can be applied to arbitrary geometry MOC solver successfully, and may obtain higher efficiency than the original characteristics solver does because of its spectacular effect on reducing both the number of outer iterations and the total computing time. Moreover, the results could be improved by Lyusternik-Wagner extrapolation technique in some cases.
New Progresses in the IFMIF Target and Test Cell Design and a Proposal for the Specimen Flow
Authors: Tian Kuo , Arbeiter Frederik , Eilert Dirk , Heinzel Volker , Heupel Tobias , Mittwollen Martin , Mo¨slang Anton , Scheel Nicola , Stratmanns Erwin
Published: April 2011

The development of the conceptual design of the IFMIF Target and Test Cell (TTC) is briefly summarized by outlining the previous reference TTC design and the current Modular TTC (MTC) concept. Based on the MTC concept, the latest progresses of the preliminary engineering design of the key TTC components, including the TTC vessel, the Top Shielding Plugs (TSPs), the Removable Intermediate Ring (RIR), and the Test Module Interface Head (TMIH), are described. A specimen flow, based on handling requirements of the High Flux Test Module (HFTM), between the TTC, Access Cell (AC), Test Module Handling Cell (TMHC) and the Post Irradiation Examination (PIE) facilities is proposed as well as the function of the AC and TMHC is preliminary defined. The TMHC is proposed to be divided into a Component Handling Cell (CHC) and a Rig Handling Cell (RHC) regarding the dimension differences of the components to be handled inside of the cells. The recycling of the irradiated specimens for another campaign of irradiation is also considered in this specimen flow.
EPR™: A Reactor to Support Economical Fuel Cycles
Authors: Zheng S. , Meinl R. , Stephens J.
Published: April 2011

The EPR™ reactor has been designed by AREVA to support economical fuel cycles. The progress in the reactor and systems design improves the reactor safety, and allows the EPRTM reactor to support the large range of high performance fuel management strategies covering cycle length from 12 to 24 months. Different fuel management strategies with 12, 18 and 24 month cycles are described. Economic analyses are performed to illustrate the low uranium consumption and the high fuel cycle performance compared with the fuel managements implemented in most current traditional PWR reactors.
Development and Demonstration of a New Testing Capability for Simulating Multiple Fuel Assembly Impact During a Seismic Event
Authors: Zhao W. , Lee Y. C. , Kunishi H. , Karoutas Z. , Doan A.
Published: April 2011

As an important part in both product development and analytical model validation, Westinghouse continues to develop prototype testing capabilities. Towards better understanding fuel assembly’s structural performance during a seismic event, our recent effort was to develop a new testing capability for simulating multiple fuel assembly impact that may occur during a seismic event. The test system was developed by adding a pendulum system to our existing fuel assembly mechanical test stand. Following qualification of the testing system, a prototype fuel assembly was tested using selected pendulum parameters. Fuel assembly displacement, impact loads at various contact locations, and the spacer grid deformation as a function of impact forces were obtained. The test results provided new insights and understanding of the fuel assembly’s structural performance during a simulated seismic event, and are useful for model development and validation.
Steam Generator Asset Management Program
Authors: Prabhu Padmanabha J. , Testa Damian A.
Published: April 2011

The Steam Generator Asset Management Program (SGAMP) is a long term program designed to maximize the performance and reliability of the steam generators. The SGAMP focuses on plant specific conditions and hence is applicable to the original or the replacement steam generators. It is recommended that the utility and the vendor form a joint steam generator management team (SGMT) to develop, monitor and implement a long-term plan to address steam generator operation, maintenance and life extension goals. The SGMT will consist of representatives from operations, chemistry, maintenance and engineering functions and will be responsible for making decisions related to the steam generators. The charter of the SGMT is to develop a steam generator strategic plan that will cost-effectively manage steam generator options. The strategic plan is consistent with the Steam Generator Program Guidelines (NEI 97-06 in the United States). The strategic plan is a living document and is revised periodically to incorporate inspection results, new technology developments, lessons learned and industry experience. Cost-benefit analyses of strategies may be performed to prolong steam generator operability through steam generator performance modeling (tube degradation, fouling, etc.), diagnostic tools, regulatory strategy, condition monitoring and operational assessment strategy, and maintenance strategy. The SGMT will provide input regarding potential maintenance of the steam generators with schedule and cost impacts for each outage. It will also recommend engineering evaluations to be performed in support of program goals and will develop short- and long-term recommendations. These recommendations will address action plans, performance measures and results. Secondary side inspection and cleaning strategy should be developed (techniques and frequency) to maximize performance cost-effectively. This paper is based on Westinghouse experience gained by working with several pressurized water reactor (PWR) plant operators in the United States (US).
Approaching Dynamic PSA Within CANDU 6 NPPs
Authors: Lontos M. , Mazzini M. , Mazzini D.
Published: April 2011

The plant dynamics and the interactions of plant systems and operator with the random evolution of parameters led to the development of dynamic reliability methodologies. The present paper approaches a dynamic PSA methodology by making use of the thermal–hydraulic model of CANDU 6 reactor, implemented in RELAP5 Mod.3.3, and of IDDA code. IDDA (Integrated Dynamic Decision Analysis) is a software code that with the help of enhanced dynamic event tree methodology provides results in terms of system unavailability. The coupling of the two above mentioned codes allows a full representation of the plant operational states, as well as of all the possible occurrence patterns that complete the spectrum of possible probability-consequence conditions. The plant transient considered for the CANDU 6 thermal–hydraulic model is the total Loss of Feed Water (LOFW) supply to the secondary side of steam generators. That is followed by depletion of water inventory and subsequent cool down via Emergency Water System (EWS). The present dynamic PSA approach reveals those situations where the correct intervention of protective equipment could bring to unexpected events. This allows taking the most appropriate decisions for the given plant configuration.
Undesirable Flow Behavior in a Proposed Validation Data Set
Authors: Johnson Richard W. , McIlroy Hugh M. , Johnson Ryan C. , Christensen Daniel P.
Published: April 2011

The next generation nuclear plant (NGNP), whose development is supported by the U. S. Department of Energy, will be a very high temperature reactor (VHTR). The VHTR is a single-phase helium-cooled reactor that will provide helium at up to 800 °C. The prospect of a coolant at these temperatures circulating in the reactor vessel demands that careful analysis be performed to ensure that excessively hot spots are not created and that sufficient mixing of the coolant is obtained. Computational fluid dynamics (CFD) coupled with heat transfer will be used to perform the desired analyses. However, primarily because of the imperfect nature of modeling turbulent flow, any CFD calculations used to perform nuclear reactor safety analysis must be validated against experimental data. Experimental data have been taken in a scaled section of the lower plenum of a prismatic VHTR at the matched index of refraction (MIR) facility at the Idaho National Laboratory. These data were taken with the intent that they be examined for use as validation data. A series of investigations have been conducted to assess the MIR data. Issues that have already been examined include the extent of the required computational domain, the outlet boundary condition, the inlet data and the effect of the turbulence model. One of the jets that flow into the model impacts on a wedge, which represents a portion of a hexagonal graphite block that is part of the inner wall of the lower plenum. The nature of the flow below this particular jet is such that a randomly varying recirculation zone is created. This recirculation zone is seen to change in size, causing a relatively long-time scale of motion or disturbance on the flow downstream. It is concluded that such a feature is undesirable in a validation data set, firstly because of its apparent random nature and, secondly, because to obtain an appropriate longtime average would be impractical because of the compute time required. It is found that by eliminating the first of the four inlet jets into the scaled model, the resulting recirculation zone is rendered stable.
Resonance Calculation Code UFOP Based on the Hyper-Fine Group Neutron Resonance Calculation Method
Authors: Qin Yulong , Wu Hongchun , Cao Liangzhi , Liu Qingjie
Published: April 2011

Resonance self-shielding calculation is very important in reactor physics calculation. Conventional resonance calculation method has some fundamental defects, which hinders its application in some problems. The Hyperfine Energy Group Resonance Calculation Method is studied in this paper and a code named UFOP is developed based on this method. In this method, the resonance energy range is divided into hyperfine energy intervals (tens of thousands) and the collision probabilities are calculated. Then the slowing-down equation is directly solved based on CPM (collision probability method). Some techniques are applied in solving the slowing-down equation for improving computational efficiency and reducing calculation error. A resonance benchmark problem with homogeneous and infinite material is calculated to validate the accuracy of the computation code and the hyper-fine group cross-section library utilized in the code. A PWR fuel cell is also calculated and the results are compared with MCNP. The results show good accuracy of this method and the validity of UFOP code.
The Thermal-Hydraulics Behavior Analysis Before the CANDU Fuel Channel Dry-Out
Authors: Yuan Jingtian , Tong Lili , Cao Xuewu
Published: April 2011

The CANDU 6 nuclear power plant is modeled with SCDAP/RELAP5 code including the heat transport system (HTS), pressure and inventory control system, calandria vessel (CV) and main steam lines, etc. Through simulation of station blackout (SBO), the results obtained before the fuel channel dry-out are illustrated. Before the steam generators (SGs) dry out, the core decay heat could be removed through the SG secondary side, and fuel temperature, pressure of the HTS could keep low. The loss of the HTS inventory results in the fuel channel dry-out, heat-up of the fuel, the calandria and pressure tubes. The moderator temperature remains low in the whole process.
Numerical Research on Complex Convection in the Fuel Transfer Channel
Authors: Guo Qiang
Published: April 2011

An investigation is made on the thermal transfer characteristic of horizontal convection in the fuel transfer channel. The convection is a complex 3-D flow. Influencing factors on the convection are calculated and compared, such as the stuck position of fuel assembly, the diameter of channel, and the mechanical structure of the fuel transfer system. An optimized mechanical design for fuel transfer system is suggested. It seems effectively improved cooling effect by numerical simulation validation.
Prediction of Pressure Drop in Air-Water Two Phase Upflow Through Packed Bed
Authors: Zhang Nan , Sun Zhongning
Published: April 2011

The resistance characteristics of air-water flow upward through packed bed have been studied experimentally. Experiments were conducted in transparent tube with 50mm inner diameter filled with glass spheres of which the diameters are 2, 5 and 8mm, respectively. Experimental results show that the pressure drop increases with the increasing of gas-liquid mass flow rate, and has a certain relationship with the flow pattern. The different particle diameter and porosity have great influence on pressure drop under the same flow condition. The applicability of several representative correlations for calculating the two-phase pressure drop and two new correlations were evaluated against 234 group experimental data. All compared correlations can be grouped into two, namely: (a) the separated flow model based on Lockhart & Matinelli method (b) the homogeneous model based on gas and liquid Reynolds number. The results show that: (1) the best agreement between measured and calculated values is obtained with the correlations based on separate flow model, but the predictive ability is reduced with the increasing of particle diameter. (2) the existing homogeneous model shows the considerable discrepancies with experiment values; however the modified homogeneous correlation by considering particle-to-column diameter ratio and porosity gives good agreement with experimental data.
Research of Radon Detection Based on Air Mesh and Pulse Ionization Chamber at Normal
Authors: Fang Fang , Dan Li , Zhang Haiwei , Wei Chen , Wei Zhou , Huang Hongquan
Published: April 2011

The research of radon detection based on air mesh and pulse chamber at normal is presented in details. The Instrument consists of air mesh, pulse chamber, signal amplification and discriminator and MCU. Radon of ambient air can directly make a free diffusion to the high-pressure chamber, because The Instrument uses a new type of atmospheric air-gridded pulse ionization chamber, ventilated with ambient air. A voltage pulse in the Resistance connected to Collecting electrode is produced when the charged particles moving through the gas in high-pressure chamber, can make the air ionize to create the charged ion-pairs (negatively charged ion and positively charged ion), which drift to the cathode and anode respectively with the effect of the directional electric field. The voltage pulse height represents the ionizability of an incident particle. And the relative counting can be made for the incident particles, reflecting the changes of the radon concentrations in air, and the total measurement too. But the pulse signal, from Front-end detector, is very weak, so they need to be amplified and shaped by amplification and discriminator circuit and RC-CR band-pass filter, compared with a given threshold voltage through the Pulse discriminator to produce the digital signals which micro-controller system can handle and for which SCM can make the counting, storage and display. Then, this Instrument is a design of appropriate devices for short-term measurements, and some criteria used in the design of this instrument are field measurements applicability, portability, convenience and reliability.
Thermal Interaction of Lead-Alloy Droplet With Subcooled Water in Pool Water Tank
Authors: Sa Rongyuan , Takahashi Minoru
Published: April 2011

Visualization test was conducted by designing and setting up a special experimental apparatus to investigate characteristics of thermal-hydraulic interaction between lead-alloy droplet and subcooled water in pool water tank. The violent boiling phenomena were observed by a high-speed camera and local transient pressure was measured by high frequency piezo pressure transducer. The results showed that violent boiling mainly occurred when interface temperature between droplet and water was higher than homogenous nucleation temperature of water and subcooling temperature of water was higher than 40K. A violent boiling model for lead-alloy droplet/water interaction was proposed. In the model, partial contact of lead-alloy droplet with water caused by vapor film instability was taken into account to simulate fragmentation due to rapid evaporation.
Flow-Induced Void Fraction Transition Phenomenon in Two-Phase Flow
Authors: Shen Xiuzhong , Mishima Kaichiro , Nakamura Hideo
Published: April 2011

The flow-induced void fraction transition phenomenon was observed in an upward air-water two-phase flow in a vertical pipe with inner diameter D = 200 mm and height z = 25 m. As the two-phase flow develops in a vertical pipe, the void fraction increases firstly in the flow direction in bubbly flow, then decreases in the flow direction, finally increase again. The flow-induced void fraction transition shows an N-shaped changing manner. The present experimental investigation revealed that this phenomenon was attributed to the formation and the growth of local dominant large bubbles in the flow. According to the bubble sizes and behaviors observed in the experiment, the flow regimes were classified into bubbly, churn and slug flows in a vertical large-diameter pipe. The drift velocities in the three flow regimes were measured in this paper. New constitutive equation for drift velocities in bubbly, churn and slug flows was proposed and confirmed in this study. The flow-induced void fraction transition in N-shaped manner can be predicted by using the drift flux model with the newly developed constitutive equations.
Solving the Time-Dependent Transport Equation Using Time-Dependent Method of Characteristics and Rosenbrock Method
Authors: Yang Xue , Jevremovic Tatjana
Published: April 2011

A new approach based on the method of characteristics (MOC) and Rosenbrock method is developed to solve the time-dependent transport equation in one-dimensional (1D) geometry without any approximation and considering delayed neutrons. Within the MOC methodology, the leakage term in time-dependent transport equation can be simplified to spatial derivative of the angular flux along the characteristics lines. For 1D geometry, the proposed exponential correlation derived from the steady-state MOC equations provides the correlation between the cell outgoing angular flux and the cell average angular flux. Thus, the spatial derivative term can be further substituted by the relation containing only the cell average angular flux that represents the unknowns. Therefore, the 1D time-dependent transport equation is decomposed into a series of locally coupled ordinary differential equations (ODE). Rosenbrock method was chosen to solve the system of ODEs. It is a fourth order explicit method with automatic step size control feature developed for stiff ODEs. The FORTRAN90 numerical program is developed to thus solve the time-dependent transport equation considering delayed neutrons in 1D geometry with both vacuum and reflective boundary conditions. The step perturbation is currently supported. The method presented in this paper was verified in comparison to 1D fast reactor benchmark showing good accuracy and efficiency.
Safety Analyses of the Lead-Bismuth Eutectic Cooled Accelerator Driven System XT-ADS
Authors: Chen Xue-Nong , D’Andrea Danilo , Matzerath Boccaccini Claudia , Maschek Werner
Published: April 2011

Safety analyses for the XT-ADS were performed with the reactor safety code SIMMER-III. Besides a brief description of the numerical model, three typical transients are presented in this paper, namely, the unprotected loss of flow (ULOF), unprotected transient over-current (UTOC), and the unprotected coolant flow blockage accident (UBA). Because of the important phenomenon of mass flow rate undershooting in the ULOF case, an integral equation model was set up for a further theoretical study of ULOF. The model confirms the numerical simulation results for various cases and gives a deeper understanding of this phenomenon. The faster the pump shut down, the larger is the undershooting of the mass flow rate. On the other hand a larger coolant cold leg area leads to a weaker undershooting. The stability analysis shows that the natural convection state is in the region of the damped oscillation for the current XT-ADS design.
Experimental Research of Pebble Dynamics of Two-Region Reactor Core in HTR
Authors: Zhang Xiuzhi , Zhang Youjie , Yang Xingtuan , Liu Zhiyong , Jiang Shengyao
Published: April 2011

According to the similarity law, the experimental system of pebble dynamics of two dimensions reactor core in HTR can be established. With the execution of phenomenological research of the moving rules of pebbles, it is available to obtain distribution of path lines, curves of degree of dispersion, and distribution of stagnant time. The results turn out to be that (1) the motion of a single pebble is partially diffused and random in micro view, but the rule of all pebbles is assured and steady in macro view; (2) curves of degree of dispersion continue to increases at first and then decrease gradually. The degree of dispersion of bilateral pebbles is bigger than that of central ones; (3) stagnant time is mainly linear to height and the ratio of stagnant time of radical location and central location increase when a pebble is moving down.
Transmission Type Ultrasonic Imaging of Horizontal Two-Phase Flow
Authors: Chang Jen-Shih
Published: April 2011

Ultrasonic technique was widely applied for the measurement of two-phase flow, however, the majority techniques deal primary with pulse echo technique under higher operating sound frequency in a MHz order since higher accuracy can be obtain for the dynamic gas-liquid interface locations. On the other hand the transmission type ultrasonic methods was also developed for bubbly flow diagnostics for the time and cross sectional averaged void fraction measurement or bubble two-phase flow parameter determination by pulse echo techniques with higher operating sound frequency in a MHz order. In this work, the transmission type ultrasonic imaging of horizontal two-phase flow was experimentally investigated. Ultrasonic transducer used is 150 kHz and time averaged void fraction results for stratified flow was compared with capacitance void measurement and ultrasonic pulsed echo techniques. The results show that the transmission intensity for the stratified flow pattern decreases with increasing void fraction until 55% of void then increases with increasing void fraction till 100%. An application to plug, slug and annular flow regime also investigated in detail.
Experimental and Numerical Simulation of Radiolysis Gas Detonations and Mechanical Response of BWR Exhaust Pipes
Authors: Kuznetsov M. , Redlinger R. , Breitung W. , Grune J. , Sempert K. , Friedrich A.
Published: April 2011

Radiolysis gas (2H2+O2) can accumulate in BWR steam piping in case of steam condensation. An ensuing detonation of the radiolysis gas is the likeliest cause of a pipe and/or valve damage. In the current work we investigate a typical BWR exhaust pipe, which connects the high pressure steam piping with the ambient atmosphere, under the following “worst case” scenario: (a) accumulation of radiolysis gas in an exhaust pipe, (b) fast valve opening to the high pressure system with steam at 70 bar, and (c) adiabatic pressurization of the radiolysis gas by the steam. Taking into account a water surface level of 6 m from the open end this leads to an equilibrium state of 20 bar pressure and 602 K temperature for the pressurized radiolysis gas. The main purpose of the current work was an experimental and numerical evaluation of the maximum pressure load and the integrity of the BWR exhaust pipe in case of a detonation of the pressurized radiolysis gas.
Digital Image Evaluation Method for Digital Radiography
Authors: Mori Masako , Kashiwa Toshibumi , Aoki Yoshimitsu
Published: April 2011

Digital radiography is getting one of the common radiographic testing techniques in various industries now. However, to apply this new technique to nuclear components radiographic testing, one big issue is how we can evaluate and ensure that the taken images have enough image qualities to be used as inspection record. In film radiography, the IQI, which stands for Image Quality Indicator, have been used to ensure that taken films have enough quality to detect any specified defects in the products. So in this paper, new alternative IQI that developed in our previous study for digital radiography to evaluate digital image quality are tested and evaluated. In addition, image evaluation criteria are also developed and evaluated by calculating MTF, which stands for Modulation Transfer Function, from the IQI images taken with the products. Finally, the recommended procedures to evaluate radiographic testing image are summarized.
Dynamic Analysis of Including New Loads on Spent Fuel Storage Rack
Authors: Tsai YaoLong , Tsai TaiPing , Li ChaoJen , Wang Li-Hua , Fan TeWei , Chou ChangPin
Published: April 2011

The safety evaluation of spent fuel storage rack is performed for many power companies on the seismic loading previously. However, due to the inclusion of new loads. i.e., Safety/Relief Valve Actuation (SRV) and Loss-of-Coolant Accident (LOCA), the reevaluation is required according to the regulations of NRC and AEC. Instead of simplifying one rack as one beam with many degrees of freedom, all racks in the pool are discretized by the shell elements in ANSYS. In addition, hydrodynamic mass model between racks and between fuel and rack are included.