bluuebunny/crossref_metadata_2025 huggingface dataset:
Full Digital I&C System in Tianwan Nuclear Power Station: Summary of Operational Experience and Analysis of Typical Events
Authors: Shi Qingwei , Zhou Ping
Published: April 2011

The paper introduces the first full digital Instrumentation & Control System (referred to as “I&C System” hereinafter) in Tianwan Nuclear Power Station in China, in which the computer and Ethernet network were adopted as the technique platform. After introduction of the characteristics and composing of the I&C system, some problems appeared during commissioning and normal operation are described, mainly focuses on I&C system design optimization, major defects elimination and solution, reasons analysis and corrective remedies for typical events. Since the two units were in commercial operation in 2007, “nuclear safety culture” of the station has been promoted persistently, management and administration regulations are strict and precise, responsibilities are clearly assigned, and documentation system is completely developed, work is being continuously improved. The full digital I&C system is maintained by qualified I&C staffs strictly in accordance with the procedures; The full digital I&C system has exhibited its excellent qualities and remarkable characteristics as follows: stable operation, reliable function, friendly man-machine interface, intuitionistic operation and monitoring, easy configuration, and convenient trouble-shooting. Tianwan Nuclear Power Station I&C system has never caused any event that led to change of the unit operation status, thus has contributed great deal to the safe, stable and economic operation of the two units.
Design of the Moderator and Cryogenic System for Generating Cold-Neutrons at the CPHS
Authors: Feng Qixi , Feng Quanke , Zhong Bin , Liang Tianjiao , Kawai Takeshi , Wei Jie , Loong C.-K.
Published: April 2011

The Compact Pulsed Hadron Source (CPHS) of Tsinghua University will produce neutrons by the Be(n,p) reaction through bombarding a proton beam with 13MeV/50Hz/1.25mA from a LINAC system on a beryllium target. One of the purposes of this neutron source facility is to provide the neutron scattering capability for characterization of materials, especially soft matters and biological systems. Cold neutrons (wavelength > 4 Å) are essential to characterize the structure of these materials over the length scale of ∼100 nm with good resolution. We discuss the design and optimization of a cold neutron source (CNS) which employs a solid methane moderator for cold neutron generation. The moderator configuration, the associated cryogenic system, and operation conditions will be discussed.
Visualization of Bubbles Behavior in Lead-Bismuth Eutectic by Gamma-Ray
Authors: Takahashi Minoru , Yumura Takanori , Yoda Isao , Sa Rongyuan
Published: April 2011

As a fundamental study for a direct contact type of lead-bismuth-cooled fast reactor (PBWFR), the method of gamma-ray radiography was applied to the visualization of the nitrogen bubbles in lead-bismuth eutectic (LBE) in a vertical square duct with inner size of 10×10×73.3 mm3. Nitrogen gas was injected from a bottom nozzle with inner diameter of 3mm at the flow rate from 0.13 to 0.40NL/min. As γ–ray source, 60Co with 11TBq was used, and a multi-crystal Gd2O2S (Tb) scintillator was employed for detection of γ–ray. The bubble behavior was visualized successfully by means of this method. Void fraction distributions were obtained by processing the visualized images, and the bubble length and bubbles rising velocity were determined from the result. There were some unstable bubble behavior depending on gas injection flow rate.
Study on SCWR Fuel Assembly Design With Thermal-Hydraulic and Neutron-Physical Characteristics
Authors: Li Zhenyang , Zhou Tao , Sun Canhui , Liu Xiaozhuang
Published: April 2011

Physical characteristics of the coolant in the Supercritical-pressure Light Water Cooled Reactor (SCWR) vary greatly near the pseudo-critical point, which will cause variations of core neutron cross section and then bring about power perturbation. Further it will prompt the corresponding thermal parameters of supercritical water changed, and form feedback action, finally resulting in intensely coupled thermal-hydraulics and neutron-physical. Proper fuel assembly has been selected as research object, and the model of multiple parallel channels has been established. In view of this model, using DRAGON code for neutron-physical calculations and developing corresponding thermal-hydraulic programs, and then achieve coupling them through appropriate data interface, the calculation platform established. Finally the power distribution and the corresponding parameters temperature distributions in the model have been predicted. On account of deficiencies reflected in calculations, such as the heterogeneous power distribution, fuel assembly geometry has been changed, for instance the proper peripheral moderator wall has been added based on the preceding assembly, then do the coupling calculations and analyze the results. Comparisons between different results have been made, and the expected aim has been reached, which can prove the rationality of assembly modifications and meanwhile prove the usability of the calculation platform. Thus, modified assembly and the calculation platform could be the calculation foundation in future designs of SCWR.
Numerical Simulation of Fluid Flow and Heat Transfer Under Flow Fluctuation Condition in Horizontal Tube
Authors: Liu Yusheng , Gao Puzhen , Xing Dianchuan
Published: April 2011

Fluctuating flow is widely presented in nuclear power plant operating procedure. When the fluctuating flow occurs in the loop, the fluid flow and heat transfer in the core will be affected, which makes the study of flow fluctuation have more practical significance. With computational fluid dynamics (CFD), characteristics of fluid flow and heat transfer are numerically simulated in a horizontal tube under periodical fluctuating flow. The influences of different factors on the fluid flow and heat transfer are analyzed. The simulation results of steady flow and heat transfer in horizontal tube agree with the traditional empirical correlations’ results, which validates the feasibility of doing this research using CFD simulation. The horizontal tube fluctuation flow and heat transfer with different flow fluctuation periods, fluctuation relative amplitudes and heat fluxes are numerically simulated. The results show that the smaller the flow fluctuation period is, the larger the flow fluctuation relative amplitude we get, and the more evident influence of flow fluctuation on fluid flow and heat transfer can be found. The larger the heat flux is, the larger amplitude of temperature fluctuation of fluid will be. What is more, there is a lag in phase between friction coefficient and velocity, which is not presented between heat transfer coefficient and velocity.
Design of Wireless Heterogeneous Framework for Radiation Monitoring in Nuclear Power Plant
Authors: Gao Shouwei , Gu Danying , Zhang Weidong , Zhu Xiaojin
Published: April 2011

Radiation monitoring plays a vital role in the safe and efficient operation of the nuclear power plant (NPP). The current radiation monitoring system (RMS) generally uses cable monitoring network with distributed radiation monitors. It will introduce various compatibility issues that more detector nodes are added to the existing cable monitoring network. The communication protocols from different device manufacturers are not compatible. Furthermore, the original RMS has to be shut down for rewiring and reconstruction. In this paper, a heterogeneous framework is proposed based on the wireless sensor network (WSN) technology, for monitoring environmental conditions around and inside NPP, specifically, radiation levels. The proposed full-scope RMS has a no-wiring and no-construction upgraded scheme based on the WSNs, which forms a heterogeneous multi-networks fusion control system, and does not affect the existing NPP radiation monitoring facilities. The introduction of the wireless gateway to build heterogeneous monitoring framework makes it possible to complete the system seamless upgrade with a lower cost and higher feasibility.
A Study on Evaluation Method About Human-Machine System Interface for Nuclear Power Plant
Authors: Sun Yongbin , Gu Pengfei , Jiang Guojin
Published: April 2011

Now the new control technology has been used in industry including the main control room design of nuclear power plant (NPP). Although We have designed the computerized human-machine system interface (HSI) in Ling Ao Phase II NPP, the old evaluation method about HSI can not be used for the new one. Based on this experience together with reference to the other published information such as NUREG-0700 Rev.2, this study have developed an evaluation method based on a model of human cognitive processes in order to provide the viewpoint of the evaluation on the operability of interface. The method describes systematically the human error categories and evaluation items for HSI. And the method has been evaluated using the evaluation of experimental results obtained by a simulator equipped with computers and touch operation control panels.
Collapse Limit of Water Jetting With Surface Condensation in Steam Injector
Authors: Ikuta Ryuhei , Koizumi Yasuo , Yoshida Hiroyuki
Published: April 2011

Collapse of a water jet flowing out from a nozzle to the atmosphere was examined. The diameters of nozzles used in the experiments were 3, 6 and 8 mm. The flow state of the water jet was recorded with a high speed video camera. The collapse length was derived from recorded images. When the flow velocity was quite low, the surface of the water jet was smooth and small perturbations appeared at the lower position of the water jet. As the flow velocity was increased, the position where the small perturbations appeared came close to the nozzle outlet. The perturbations grew as these went downstream and lumps of water were formed at the lower position. When the flow velocity was further increased, successive waves came around on the surface of the water jet. The collapse of the water jet occurred in such a state that the lump of water was torn off from the jet. When the water jet velocity was high, the jet turned into a dispersed flow and the collapse occurred. The agreement of the measured results and the predicted results was poor. It was considered that the instability of the surface of the water jet seemed important for the jet collapse in the present experimental range. The Kelvin-Helmholtz instability wave length was compared with the measured wave length on the water jet. When the wave length reached the Kelvin-Helmholtz instability wave length, the jet collapse occurred except the case that the transition to the dispersed flow caused the jet collapse.
Direct Observation and Control of Liquid Sodium Flow Dynamics Using VUV-LIF-PIV Technique Under EXB Lorentz Force
Authors: Fukuda Takeshi , Takata Takashi , Horiike Hiroshi , Kimura Nobuyuki , Kamide Hideaki
Published: April 2011

Based on a heuristic hypothesis that alkaline metals with one single electron on the outermost shell would little interact with externally applied RF field above plasma frequency, a rudimentary experiment as well as the theoretical estimate of the energy structure been performed to further explore the spectroscopic properties of liquid sodium (Na)[1–3]. Consequently, it was successfully proven that Na is reasonably transparent to the VUV (vacuum ultraviolet) laser radiation, although the liquid Na surface is highly reflective, being like a mirror to human eyes. The impact of this result is that the velocity field information inside the liquid Na can be visualized by implementing the well developed PIV (particle image velocimetry) technique[4, 5]. A large eddy simulation (LES) code has also been developed for comparison with the experimental results[6]. Furthermore, the newly developed Na loop is designed so as to enable the application of electric and magnetic field in the orthogonal direction to each other that vigorous dynamics of vortices inside the liquid Na are resolved in the phase space under the Lorentz force. The results herein obtained contribute not only to the thermal hydraulics in fast reactors[7, 8] but also space physics, such as the spiral galaxy formation[9] and solar flare activities[10].
Solid Particle Resuspension Model Development for the GASFLOW Code
Authors: Xu Zhanjie , Travis John R. , Jordan Thomas
Published: April 2011

Safety reports have shown that tons of solid particles would be generated as dusts in the operation of ITER facility. The dust particles include carbon, beryllium and tungsten with diameters ranging from a few to a few hundreds microns. The particles deposit downwards and mostly accumulated on the surfaces of the diverter on the bottom side of the vacuum vessel (VV). In accident scenarios, e.g., loss of vacuum accident (LOVA), the potentially combustible dust particles can be suspended by the air ingress and entrained into the whole volume of the VV, and impose a risk of dust explosions in case of unintentionally ignition to the whole ITER facility. Therefore the mechanism of particle resuspension was investigated theoretically in the work. A force balance approach and numerical fittings have been utilized to develop a semiempirical particle resuspension model based on a group of particle resuspension experimental data. The model has been applied into a three-dimensional computational fluid dynamics code, GASFLOW. The model validation has been done by comparison of the numerical predictions about particle resuspension rates in given incoming flows against the corresponding experimental data. The comparisons have proved the validity of the developed model about particle resuspension.
Flow Mixing Inside a Control-Rod Guide Tube: Part I—CFD Simulations
Authors: Tinoco Hernan , Lindqvist Hans , Odemark Ylva , Ho¨gstro¨m Carl-Maikel , Angele Kristian
Published: April 2011

Two broken control rods and a large number of rods with cracks were found at the inspection carried out during the refueling outage of the twin reactors Oskarshamn 3 and Forsmark 3 in the fall of 2008. As a part of an extensive damage investigation, time dependent CFD simulations of the flow and the heat transfer in the annular region formed by the guide tube and control rod stem were carried out, [1]. The simulations together with metallurgical and structural analyses indicated that the cracks were initiated by thermal fatigue. The knowledge assembled at this stage was sufficient to permit the restart of both reactors at the end of year 2008 conditioned to that further studies to be carried out for clarifying all remaining matters. Additionally, all control rods were inserted 14% to protect the welding region of the stem. Unfortunately, this measure led to new cracks a few months later. This matter will be explained in the second part of this work, [2]. As a part of the accomplished complementary work, new CFD models were developed in conformity with the guidelines of references [3] and [4]. The new results establish the simulation requirements needed to accomplish accurate conjugate heat transfer predictions. Those requirements are much more rigorous than the ones needed for flow simulations without heat transfer. In the present case, URANS simulations, which are less resource consuming than LES simulations, seem to rather accurately describe the mixing process occurring inside the control rod guide tube. Structure mechanics analyses based on the CFD simulations show that the cracks are initiated by thermal fatigue and that their propagation and growth are probably enhanced by mechanical vibrations.
Pressure Measurement Test of Single Elbow Simulating Na Cooled Fast Reactor Cold-Leg Piping
Authors: Ebara Shinji , Aoya Yuta , Sato Tsukasa , Hashizume Hidetoshi , Yuki Kazuhisa , Aizawa Kosuke , Yamano Hidemasa
Published: April 2011

Regarding the Japan Sodium-cooled Fast Reactor, a multi-elbow piping system is adopted for its cold-legs. Flow Induced Vibration (FIV) is considered to be caused by complex flow with very high velocity in the elbows. In this study, pressure measurement test of a single elbow flow is conducted to find out pressure fluctuation characteristic which is related to the elbow turbulent flow and lead potentially to the FIV. Two types of experimental loops, that is, 1/7 and 1/15-scale setup simulating the JSFR cold-leg pipings, are used for pressure measurement, and a distinguishing peak can be seen in the power spectrum density profile of pressure fluctuation obtained where flow separation occurs and at the downstream from it. This characteristics of pressure fluctuation is obtained from the two different scale experiments, and the scale effect is not found in terms of the pressure fluctuation.
Aspects of Minimizing Steel Corrosion in Liquid Lead-Alloys by Addition of Oxygen
Authors: Schroer Carsten , Wedemeyer Olaf , Konys Juergen
Published: April 2011

The concept of minimizing steel corrosion in liquid lead-alloys by addition of oxygen strongly depends on the availability of efficient devices for oxygen transfer and reliable oxygen sensors. The accuracy of electrochemical oxygen sensors is analyzed on the basis of theoretical considerations and results from experiments in stagnant lead-bismuth eutectic (LBE). Additionally, the feasibility of gas/liquid oxygen-transfer and the long-term performance of electrochemical sensors in flowing liquid metal are addressed based on operation of the CORRIDA loop, a facility for testing steels in flowing LBE.
The Optimization of Mathematic Model of Penumbral Imaging System’s Point Spread Function
Authors: Jiang Ren-qiang , Wu Yue-lei , Lu Zhi-rong , Zhang Tie-jun
Published: April 2011

In inertial confinement fusion (ICF) project, there are many diagnostic methods. Neutron penumbral imaging is one of the important technologies to diagnose the information about neutron spatial and temporal distribution in burn region of the core of a compressed pellet in the low yield fusion. In this study, a linear space invariant neutron penumbral imaging system was designed and established with Monte Carlo method, the system’s point spread function (PSF) was obtained. By fitting the obtained PSF, several mathematic models were obtained and compared. The improved “logistic function” mathematic model was chosen to reconstruct the coded penumbral image and the original neutron source image “T” was successfully was obtained.
LOCA Event Tree Analysis for Qinshan Phase III CANDU-6 PHWR NPP
Authors: Zheng Limin
Published: April 2011

Qinshan Phase III is the first commercial pressurized heavy water reactor (PHWR) NPP in China, and it uses CANDU-6 design developed by AECL. Based on plant design and operation experience, the event tree analysis model has been developed for both small break LOCA (SB-LOCA) and large break LOCA (LB-LOCA), which is an important aspect of operational Probabilistic Safety Assessment (PSA). Both SB-LOCA and LB-LOCA event tree analysis have been performed for Qinshan Phase III CANDU-6 PHWR NPP (TQNPC). And the event sequence development and plant damage status (PDS) were provided in the analysis. It reflects actual plant configuration and response under a certain event, and various break type and locations were also considered in the event tree analysis, e.g. Pressure Tube Rupture, Pressure Tube and Calandria Tube Rupture, Feeder Breaks, Pressurizer Relief/Steam Bleed Valves Fail Open, Liquid Relief Valves Fail Open, etc.
Thermal-Hydraulic Modeling of CANDU-SCWR and Linear Dynamic Model Development
Authors: Sun Peiwei , Jiang Jin
Published: April 2011

CANDU-SCWR is one Generation IV reactor being developed in Canada. Significant amount of efforts has been made to develop CANDU-SCWR. Little work has been done on the dynamic analysis and control design. To observe the dynamic behaviours of CANDU-SCWR, the detailed CANDU-SCWR thermal-hydraulic model is developed. The movable boundary method is adopted for CANDU-SCWR thermal-hydraulic modeling. The benefits of adopting movable boundary are derived from the comparisons with the fixed boundary method. The steady-state results agree well with the design data. The responses of CANDU-SCWR reactor to different disturbances are simulated and analyzed and the results are reasonable in theory. Linear dynamic models are derived from simulation data of CANDU-SCWR thermal-hydraulic model around the design operating point using a system identification technique to facilitate the control system design. The linear dynamic models are validated and it is shown that they can describe the dynamic characteristics of CANDU-SCWR around the design point accurately.
Analysis of Central Sloshing Experiment Using Smoothed Particle Hydrodynamics (SPH) Method
Authors: Vorobyev Alexander , Kriventsev Vladimir , Maschek Werner
Published: April 2011

Liquid sloshing phenomenon is encountered whenever a liquid in a container has an unrestrained surface and can be excited. Specific type of sloshing motion can occur during the core meltdown of a liquid metal reactor (LMR) and can lead to a compaction of the fuel in the center of the core and to energetic nuclear power excursions. This phenomenon was studied in series of “centralized sloshing” experiments with a central water column collapsing inside the surrounding cylindrical tank. These experiments provide data for a benchmark exercise for accident analysis codes. To simulate “centralized sloshing” phenomenon a numerical method should be capable to predict motion of free surface of liquid, wave propagation and reflection from the walls. A meshless method based on Smoothed Particle Hydrodynamics (SPH) for the simulation of 3D free surface liquid motion has been developed. Proposed method is applied to the simulation of “centralized sloshing” experiments. Simulation results are compared with the experimental results as well as with results of computations performed with 3D code SIMMER-IV which is an advanced reactor safety analysis code that implements the traditional mesh-based numerical method. In series of numerical calculations it is shown that overall motion of the liquid is in a good agreement with experimental observations. Dependence on the initial and geometrical symmetry is studied and compared with experimental data.
Integral Effect Test for a Inadvertent Opening of POSRV Using the ATLAS Facility
Authors: Cho Seok , Choi Ki-Yong , Park Hyun-Sik , Kang Kyoung-Ho , Kim Yeon-Sik , Kim Bok-Deuk , Kim Kyung-Doo , Baek Won-Pil
Published: April 2011

A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI (Korea Atomic Energy Research Institute). The reference plant of the ATLAS is a 1400 MWe-class evolutionary pressurized water reactor (PWR), the APR1400 (Advanced Power Reactor 1,400 MWe), which was developed by the Korean industry. The ATLAS has a 1/2 reduced height and a 1/288 volume scaled integral test facility with respect to the APR1400. It has a maximum power capacity of 10% of the scaled nominal core power, and it can simulate full pressure and temperature conditions of the APR1400. The ATLAS could be used to provide experimental data on design-basis accidents including the reflood phase of a large break loss of coolant accident (LBLOCA), small break LOCA (SBLOCA) scenarios including the DVI line and cold leg breaks, a steam generator tube rupture, a main steam line break, a feed line break, etc. An inadvertent opening of POSRV test (SB-POSRV-02) was carried out as one of the SBLOCA spectra. The main objectives of this experimental test were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation but also to present integral effect data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry.
Dynamic Behavior of Cracked Cantilevered Pipes Conveying Fluid
Authors: Cai Fengchun , Zang Fenggang , Ye Xianhui
Published: April 2011

In this article, the effects of the non-propagating open cracks on the dynamic behaviors of a cantilevered pipe conveying fluid are studied. The model divides the pipe into a number of segments from the crack sections and assembles all segments each by each by a rotational spring which has no mass. The stiffness of the spring is obtained through linear fracture mechanics. In order to obtain the modal functions which satisfy the boundary conditions and geometrical discontinuity conditions at the crack’s location, a simple approach is used. That is adding polynomial functions to the modal functions of the uncracked beam. The equations of motion for the cracked cantilevered pipe conveying fluid is derived based on the extended Lagrange equations for systems containing non-material volumes. Not only the virtual work done by the discharged fluid, but also that done by the fluid at the crack position due to the geometrical discontinuity conditions are considered in the present equations of motion. In this article, several numerical examples are given. The comparisons of solutions of the present equations with that of model in existence show that the present work is better. The influences of the relative depth, the position ratio of the cracks, the flow velocity on the eigenvalues are depicted.
PCI Risk Assessment Model
Authors: Paramonov Dmitry
Published: April 2011

The main objective of the reported study was developing a method for predicting PCI failure risk in PWR fuel during Condition 2 transients that would strike a balance between simplicity and accuracy, allow for straightforward implementation within a transient analysis methodology or core monitoring system, and include treatment of the most important PCI factors. The developed method has the following features: relies on readily calculated/available quantities such as nodal burn-up, local power, average cladding temperature and pressure differential across cladding at zero burnup, uses a power increase past gap closure as a failure criterion, replaces a bounding method with a best estimate one. It allows calculating failure risk for a given ramp power or establishing a ramp power corresponding to a certain failure probability, when relevant uncertainties are considered, provides for accuracy comparable to a fuel performance code in calculating PCI failure risk in ramp tests representative of Condition 2 transients, offers a way to infer the safe threshold from ramp test data.
Misunderstanding Regarding to Global Warming and Nuclear Power
Authors: Fukae Chiyokazu
Published: April 2011

The results of surveys conducted in some developed countries (the United States, Germany, and France) and in Japan showed that approximately 50% of the respondents considered nuclear power generation to be a cause of global warming. Therefore, it is important to investigate why a wide range of people lack the awareness that nuclear power generation is an effective means of preventing global warming and why approximately 50% of people think that nuclear power generation is a cause of global warming. In this research, it was investigated why people think that nuclear power generation is a cause of global warming. Factor analysis method was applied to data obtained from survey at Kansai area in Japan. Using the survey results, people’s awareness structure was analyzed to determine factors behind people’s perception that nuclear power generation is a cause of global warming and to identify ideas preventing people from recognizing nuclear power generation, which emits no carbon dioxide during power generation, as a means of preventing global warming. As a result, the misunderstanding, the thermal discharge and radioactive material etc. produced from a nuclear power plant promotes global warming, has influenced on this issue. It has become evident that behind such misunderstanding is a negative image of nuclear power. This negative image is a factor to decrease the evaluation that nuclear power is useful for preventing global warming regardless of the presence of the misunderstanding. It is believed that the negative image of nuclear power does not lead to direct association of nuclear power generation and global warming, but by the fear that the accident of the nuclear plant brings the environmental destruction, people evaluate that nuclear power generation is not effective for preventing global warming without grounds. Especially, the tendency is very strong in young people.
Advances in Pressure Tube Reactor Technology
Authors: Kupferschmidt W. C. H. , Duffey R. B. , Leung L. , Didsbury R.
Published: April 2011

We provide an update on recent advances in pressure tube reactor (PTR) technology as exemplified by progress in the continuous research and development path that has been defined. This includes numerous and fruitful strategic development partnerships, both nationally and internationally, to contribute to the major areas of technological innovation, economic competitiveness, energy security, sustainability and environmental stewardship. We discuss the major areas of advancement in R&D that utilize the inherent features and advantages of the unique PTR concepts of utilizing a low pressure, heavy water moderator with a modular channel construction.
Advances in Inverse Transport Methods
Authors: Wu Zeyun , Adams Marvin L.
Published: April 2011

We present advances in inverse transport methods and demonstrate their application to neutron tomography problems that have significant scattering. The problem we consider is inference of the material distribution in an object by detection and analysis of the radiation exiting from it. Our approach combines both deterministic and stochastic optimization methods to find a material distribution that minimizes the difference between computed and measured detector responses. The main advances are dimension-reduction schemes that we have designed to take advantage of known and postulated constraints. One key constraint is that the cross sections for a given region in the object must be the cross sections for a real material. We illustrate our approach using a neutron tomography model problem on which we impose reasonable constraints, similar to those that in practice would come from prior information or engineering judgment. This problem shows that our method is capable of generating results that are much better than those of deterministic minimization methods and dramatically more efficient than those of typical stochastic methods.
Study on Accident Sequences of Seismic PSA for BWR
Authors: Nishio Masahide , Yamashita Masahiro
Published: April 2011

The Examination Guide for Seismic Design of NPP was revised in 2006 in Japan. In response to the revised guide, utilities are required to establish the seismic design acceleration of the NPP site and evaluate the residual risk of individual NPP for earthquakes exceeding the seismic design acceleration. JNES is developing seismic PSA models of typical NPPs to support the regulatory review of utility’s residual risk evaluation report. Trial analysis of seismic PSA is performed for a typical BWR4 plant and a typical BWR5 plant. Dominant accident sequences and dominant initiating events are obtained, and sensitivity analysis is performed to evaluate the influence of analytical conditions on the core damage frequency (CDF) such as LOCA model and system mitigation effect. Analysis result shows that the CDF profile changes largely depending on the seismic intensity of the site.
Dynamic Response of a Nuclear Power Plant Subjected to External Accident Event
Authors: Lo Frano Rosa , Forasassi Giuseppe
Published: April 2011

In recent times there is a renewed worldwide interest in the development and application of advanced nuclear power plants (NPPs). Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR in Finland and France, AP1000 in China, etc.) and more are under consideration for licensing in several countries. Innovative NPPs are designed to be built with very broad siting conditions; therefore the safety aspects related to the external events might follow new scenarios and failure modes, different from those well known for the currently operated reactors. In this paper, the intent is evaluating the structural integrity of a nuclear containment system subjected to dynamic loadings due to a Design Base Earthquake and an aircraft impact (large size civilian jets or military aircrafts impact), which represent the two most relevant external accidents that should be considered and investigated as part of the basic design of a NPP in particular a III+ and IV Gens. In fact a suitable safety design of the NPP containment system (according to the international safety and design code guidelines, as NRC or IAEA ones), even if designed to meet other design goal, may represent a “built-in protection” to avoid or mitigate the effects of mentioned dynamic loadings. To the purpose a rather sophisticated numerical methodology, adopting finite element (FEM) approach, is employed for studying the overall dynamic behaviour of nuclear reactor and to determine the structural effects of the propagation of dynamic seismic as well as impulsive loads (containment structure response) up to the relevant nuclear components. Therefore representative three-dimensional FEM models of mentioned NPP containment and aircraft structures were set up, and used, in the performed analyses taking also into account the suitable materials behaviour and their related constitutive laws as well as the seismic excitation (determined according to the NRC rules). Moreover the performed analyses and the carried out response analyses of internal components, to both the ground motion and impact loads, were studied to check the considered NPP containment strength reserve in the case of the considered events. The obtained results seem to confirm the possibility to achieve an optimization of the NPP internal components.
Experimental Investigation of Reactor Cavity Cooling System (RCCS) for a Very High Temperature Reactor (VHTR)
Authors: Capone L. , Perez C. E. , Hassan Y.
Published: April 2011

The reactor cavity cooling system (RCCS) for a very high temperature reactor (VHTR) represents a very important safety feature for achieving the defense in depth of the plant. An experimental facility was built for testing the heat transfer capability and phenomenology of this last heat sink designed for ensuring the cooling down of structural material of the vessel and of the concrete walls of the vessel cavity. This small scale facility was built using some of the scaling laws in order to resemble the main heat transport features in RCCS configuration. The natural convection phenomena and radiative heat transfer inside the cavity were represented. The experimental facility represents half of the vessel and of the reactor cavity with five stand pipes for cavity cooling using water as cooling fluid. Measurements were performed heating up the vessel surface temperature to an average temperature of 300 °C that is the average value in accident scenarios. Temperature measurements of the vessel surface temperature, the outer pipes surface temperature profile and inlet and outlet temperature of the cooling water were performed. Axial and radial temperature profiles of the air in the cavity were measured using a movable rack of 24 thermocouples. The results demonstrated the natural circulation phenomena. In addition Velocity measurement of the air inside the cavity were performed using particle tracking velocimetry techniques (PTV) determining the flow regime characteristics and the coupling with the temperature profile. The experimental test matrix of various flow rates in the cooling pipes were carried out.
A Transient Analytic Method of Thermionic Reactor: TOPAZ-II
Authors: Hu Gu , Zhao Shouzhi , Ruan Keqiang
Published: April 2011

According to characteristics of TOPAZ-II reactor, a transient analytic method combined a nuclear reactor six-group point-kinetics model, a reactor core thermal-hydraulic model, a thermionic fuel element (TFE) performance model is established. Afterwards, establishment and debugging of this transient analytic code are completed. Verification results are reasonable agreement with reported American and Russian data. The code is original with the author.
Sino-French Cooperation on Publication of RCC Codes in Chinese Version
Authors: Wu Hong , Faidy Claude , Wu Xiaokang , Yuan Ding , Gong Lei
Published: April 2011

The present Chinese nuclear program is developing vastly and rapidly. The main type of nuclear units under construction currently is the improved generation II PWR (G2+). Taking into account that G2+ is designed based on French M310 type, and one of the technical bases and guidance that NNSA uses to evaluate G2+ is the French RCC codes, AFCEN has authorized CNPRI to translate RCC codes into Chinese version and publish them, in order to facilitate Chinese manufacturers to better apply RCC codes. CNPRI organizes several sub-groups to perform translation, revise, publication and coordination working groups, meanwhile AFCEN provides technical support. Translation is realized based on RCC codes English version, with some reference of French version. Translators work with professional technical background and knowledge, in case of non-clear understanding, including language and technical issues, they raise question list to AFCEN experts for clarification, afterward the answers and comments are merged accordingly into Chinese version translation, to assure the accuration, better understanding and the guidance to Chinese G2+ program.
Possibility for Gaining Nuclear Energy Without Radioactivity by Laser Driven Block Ignition of Solid Density Hydrogen Boron
Authors: Hora Henrich , Miley George H.
Published: April 2011

In addition to the matured “Laser Inertial Fusion Energy (LIFE)” with spherical compression of deuterium-tritium (DT) for a pure fusion engine or for fusion-fission-hybrid operation, a very new scheme may have now been opened by igniting the neutron-free reaction of proton-boron-11 (p-11B) using side-on block ignition. Laser pulses of several petawatt power and ps duration led to the discovery of an anomaly of interaction, if the prepulses are cut off by a factor 108 (contrast ratio) to avoid relativistic self focusing. In this case the Bobin-Chu conditions of side-on ignition of solid fusion fuel can be applied after several improvements leading to energy gains of 10,000 similar to the Nuckolls-Wood ignition with extremely intense 5 MeV electron beams. In contrast to the impossible laser-ignition of p-11B by the usual spherical compression, the side-on ignition is less than ten times only more difficult of DT ignition. This p-11B fusion produces less radioactivity per gained energy than burning coal. After encouraging success with computations based on the different nuclear cross sections, next steps are focusing on stability and transport problems.
A Technical Overview of The Japan’s Standards for Risk-Informed Decision Making
Authors: Narumiya Yoshiyuki , Hirano Mitsumasa , Hirano Masashi
Published: April 2011

In Japan, a lot of efforts have been made on severe accident study, and development and application of the probabilistic safety assessment (PSA) technique. The PSA was applied to the examination of the accident management (AM) plan in the beginning of 1990s and was performed for all the nuclear power plants (NPPs) to evaluate the effects of the AM. Furthermore, the PSA has been performed as part of periodic safety review (PSR) to review the safety of individual plant. In recent years, discussions have started to apply risk information to the safety regulation or safety related activities to improve rationality, accountability and transparency. As the technical foundations, the Nuclear Safety Commission (NSC) showed the safety goals and policy toward risk informed decision making (RIDM), and the Nuclear and Industrial Safety Agency (NISA) developed guidelines for risk informed regulation (RIR). Consensus standards have been developed in the Atomic Energy Society of Japan (AESJ), the Japan Society of Mechanical Engineers, and the Japan Electric Association. Especially, the AESJ established the Standards Committee in 1999 and has made several PSA standards. Now, there are seven PSA standards. The most important one is the seismic PSA standard. It was developed ahead of the world as a concrete manual. Lots of illustrations and useful example are included for ease of use and to make decision adequately. There were needs to develop a standard, which provides the basic requirements and specific procedures commonly applicable to respective fields of utilization of RIDM regarding changes in safety related activities. Responding to such needs, an implementation standard has been developed on use of risk information in changing the safety related activities. It stands over individual standards that will be developed in future, and shows the common and basic rules. It requires being consistent with the defense-in-depth philosophy, to maintain sufficient safety margins, and to clarify the influence to safety by comparing with some criteria. And it also requires as a final step that a comprehensive decision be made by considering various items, e.g. the defense-in-depth, safety margins, risk indices, and implementation and monitoring program. We will continue to make an effort toward RIDM and develop the standard to assess dominant risk hazards, e.g. fire risk and internal flooding risk. Moreover, it is necessary to develop the standard for individual applications in future.
Assessment of Early-Phase Accident Management Strategies in a Station Blackout Scenario for VVER-1000 Reactors
Authors: Tusheva Polina , Schaefer Frank , Reinke Nils , Weiss Frank-Peter
Published: April 2011

In recent years, many NPPs have developed and implemented severe accident management guidelines (SAMG). It is the primary objective of developing SAMG to prevent or mitigate the consequences of severe accidents by keeping the reactor pressure vessel (RPV) integrity and reducing the load to the containment. In a hypothetical Station Blackout accident all active safety systems are unavailable. Without additional measures this would lead to heating-up of the reactor core with severe core degradation. To avoid or to limit the consequences of a possible core heat up, different accident management strategies can be applied. This paper presents an assessment of early-phase accident management actions for VVER-1000 reactors. In particular Primary Side Depressurization (PSD) is investigated as a basic strategy for managing severe accidents under high pressure conditions. In addition, Secondary Side Depressurization (SSD) is also being investigated. It aims at fast reduction of the secondary pressure and feeding the steam generators’ secondary side with water from the feed water tank or from a different source. In that way, the heat removal from the primary to the secondary side can be significantly enhanced and the core heat-up at high pressure can be delayed. A number of simulations with different criteria for actuation of the PSD procedure and additional SSD were performed using the thermal-hydraulic system code ATHLET. This paper provides a detailed modelling of the reactor coolant system and the required safety systems, analysis of the thermal-hydraulic and safety parameters and description of the physical phenomena. Special attention is given to the possibilities of preventing or at least delaying an extended core heat-up depending on the availability of the operational and safety systems. The effectiveness of the applied accident management measures and the effect on the accident progression were studied in order to assess the maximum response time for operators’ intervention.
Deuteron Beam Driven Fast Ignition
Authors: Yang Xiaoling , Miley George H. , Flippo Kirk A. , Hora Henrich
Published: April 2011

Fast Ignition is recognized as the most promising approach to achieving the high energy gain target performance needed for commercial inertial confinement fusion (ICF). However, there are great difficulties related to the traditional approach of generating a relativistic electron beam and then focusing it on the hot “spark” region (hot spot) of the compressed target. One promising alternate approach that has been proposed by researchers at LLNL and LANL is the laser generation of a proton beam in an “interaction foil” to ignite the target. However, the total proton flux supplied from hydrogen adsorbed on the foil surface is too small to generate the desired proton flux threshold. In more recent studies it has been found that ions heavier than protons would provide better focusing on the hot spot if ways to efficiently generate them can be found. Here we propose to utilize a new “Deuterium Cluster ”type structure as the laser interaction foil to generate an energetic deuteron beam as the fast igniter. The ultra high density deuterium in the cluster structure promises much higher total flux for deuterons than for traditional protons. Also, deuterons will serve very important dual purposes — the deuteron deposition in the target hot spot will not only provide heating but also fuse with fuel as they slow down in the target. The resulting fusion alphas serve to provide added heating, reducing the input requirement. If the physics works as anticipated, this novel type of interaction foil can efficiently generate energetic deuterons during intense laser pulses. The massive yield of deuterons generated from our cluster material through laser acceleration, should turn out to be the most efficient way of FI of the DT fuel, and making the dream of near-term commercialization of FI fusion more achievable.
The Allocation of Disaster Prevention Facilities for Emergency Shelter
Authors: Yu Chu Jian
Published: April 2011

The emergency shelter with disaster prevention facilities provides a safe place by construction for emergency situations such as earthquake, nuclear accident etc., which can supply emergency evacuation and temporary residence for refugees. The disaster prevention facilities of emergency shelter are complete sets of measurements which ensure the daily essentials for refugees. The paper classifies the emergency shelters and puts forward their scales and service scopes; it also proposes the setting requirements of disaster prevention facilities for every kind of emergency shelter. The study is not only has advantages to the general plan and distribution of various emergency shelters, but also makes a significant sense of improving the level of disaster resistance and rescue.
Monte Carlo Simulation Compared With Classic Deterministic Solutions for Neutron Transport and Diffusion
Authors: Koreshi Zafar Ullah , Siddiq Sadaf
Published: April 2011

The Monte Carlo (MC) simulation method, known to handle complex problems which may be formidable for deterministic methods, will always require validation with classic problems that have evolved historically from deterministic methods [1–5] based on Chandrasekhar’s method in radiative transfer, Fourier transforms, Green’s functions, Weiner-Hopf method etc which are restricted to simple geometries, such as infinite or semiinfinite media, and simple scattering laws too for practical application. This work compares deterministic results with MC simulation results for neutron flux in a slab. We consider mono-energetic transport problem in an infinite medium and in a 1-D finite slab with isotropic scattering. The transport theory solutions used in infinite geometry are the Green’s function solution and the spherical harmonics (P1, P3) solutions, while for the 1-D finite slab, we refer to a transport benchmark for which an exact solution is available. For diffusion theory, we consider the Green’s function infinite geometry solution, and the exact and eigen-function numerical solution for finite geometry (1-D slab). The objective of this work is to illustrate the results from all the methods considered especially near the source and boundaries, and as a function of the scattering probability. The results are plotted for six elements that include a strong absorber, such as gadolinium, and a strong “scaterrer” such as aluminium. The present work is didactic and focuses on problems which are simple enough, yet important, to illustrate the conceptual difference and computational complexity of the deterministic and stochastic approaches.