bluuebunny/crossref_metadata_2025 huggingface dataset:
Nuclear Fuel Cycle: Which Strategy to Sustain Nuclear Renaissance?
Authors: Liberge Renaud , Arslan Marc
Published: April 2011

By proposing a credible response to the growing demand for energy while emitting no greenhouse gas, nuclear power will more than likely expand in the future, leading to increased quantities of used nuclear fuel to manage. Recycling the energy from this used fuel and an efficient waste management are key components for a sustainable development of nuclear energy. Current closed fuel cycle enables 25% in uranium savings, reduces the volume of waste by a factor 5 while its radiotoxicity is divided by a factor 10. Excellent track records of existing recycling plants participate to the competitiveness of MOX (Mixed Oxide) and ERU (Enriched Reprocessed Uranium) fuels compared to ENU (Enriched Natural Uranium) fuel. By offering a sound solution for nuclear waste management, recycling also contribute to favour nuclear acceptance. The advantages of closed fuel cycle have been demonstrated by the successful policy of recycling implemented in Europe for more than 30 years, with 35 reactors using MOX fuel with an excellent return of experience. The AREVA industrial recycling platform (La Hague and Melox plants) has treated more than 24,500 tons of used fuel and fabricated more than 1,600 tons of MOX fuel. With the new AREVA EPR™ reactor, recycling will take another step forward, enabling MOX fuel loading of up to 100%, thus offering optimized management options of recycled fuel, and giving more flexibility to its customers.
Remote Field Eddy Current Testing for SG Inspection of Fast Reactor
Authors: Kobayashi Noriyasu , Ueno Souichi , Nagai Satoshi , Ochiai Makoto , Jimbo Noboru
Published: April 2011

We confirmed defect detection performance of a remote field eddy current testing (RFECT) in order to inspect a helical-coil-type double wall tube steam generator (DWTSG) with a wire mesh layer for the new-type small fast reactor 4S (Supersafe, Small and Simple). As high sensitivity techniques, to increase an indirect magnetic field intensity, we focused attention on increasing a direct magnetic field intensity in the vicinity of an exciter coil by the use of an exciter coil with a magnetic material (flux guide). We adopted horizontal type multiple detector coils with flux guides arrayed circumferentially to enhance sensitivity of radial direction. According to the experimental results, the indirect magnetic field intensity (the voltage of detector coil in the region of indirect magnetic field) increased more than 100 times by the application of the exciter and detector coils with flux guides. Finally, we detected the pinhole defect of 1 mm in diameter and 20% of outer tube thickness in depth over the wire mesh layer by the adoption of the flux guides and horizontal type multiple detector coils. And we confirmed that the RFECT probe is also useful to detect the thinning defects.
Preliminary Study of Influence of Severe Accident Mitigation Countermeasures to In-Vessel Release of Fisson Products
Authors: Huang Gaofeng , Cao Xuewu , Li Jingxi
Published: April 2011

During the severe accident in a nuclear power plant, large amounts of fission products release with accident progression, which includes in-vessel release and ex-vessel release. Mitigation of release of fission products is the need of alleviating radiological consequence in severe accident. Mitigation countermeasures to in-vessel release of fission products are studied, including feed-bleed in primary loop, feed-bleed in secondary loop and cooling of ex-vessel. Representative high pressure melt accident of station blackout is chosen, and different entry condition of countermeasures is assumed. The results show that: (1) Feed-bleed in primary loop is an effective countermeasure to mitigate in-vessel release of fission products. With early time to implement the countermeasure, in-vessel release fraction of fission products is low. (2) Feed-bleed in secondary loop is also an effective countermeasure to mitigate in-vessel release of fission products. Low in-vessel release fraction of fission products is produced with early time of countermeasure implemented. (3) Cooling of ex-vessel is not an effective countermeasure to control in-vessel release of fission products, the in-vessel release fraction in this case is almost equal to base case that uses none countermeasure.
Preliminary Analysis of Three ULOFAs Based on the Dual-Functional Lithium Lead Test Blanket Module Configuration in ITER
Authors: Ni Weifeng , Qiu Suizheng , Su Guanghui , Tian Wenxi
Published: April 2011

Based on the structural design of the Chinese ITER Dual Functional Lithium-Lead Test Blanket Module (DFLL-TBM), Three Unprotected Loss of Flow Accidents (ULOFAs) were investigated preliminarily, assuming that the whole nuclear heat in TBM was carried away by the flowing lithium-lead (LiPb). The results show that the temperature of the first wall (FW) increases rapidly and the maximum temperature appears at the lower part of FW. In the analysis of ULOFAs, the maximum temperature might exceed the melting point of structure material steel. This event must be avoided by the fusion power shutdown system that terminates plasma burn.
State Oriented Procedure Application in Nuclear Power Plants in China
Authors: Yang Qingming
Published: April 2011

With the improvement of nuclear technology and nuclear safety, accidental operating procedure, as an important component of the defence in depth concept for nuclear power plant operation, develops in the way of being more robust and more operator-friendly, especially following the Three Miles Island accident. In China, nuclear power plants try to change from Event Oriented Procedure (EOP) to State Oriented Procedure (SOP) which is a scenario independent approach, developed by French institutes. This paper tries to discuss the principle of the state oriented approach and the composition of SOP. At the same time, comparison between SOP and EOP is made. Having understood the bases of SOP, China Nuclear Power Plants apply it from design to implementation appropriately only.
A Study on Human Reliability About Human Machine Interface in Accident Situation of Nuclear Power Plant
Authors: Gu Pengfei , Wang Zhifang , Zhang Jianbo
Published: April 2011

Recently Human Reliability Analysis (HRA) is becoming more important to the safety of nuclear power plant (NPP). As the reliability of the NPP equipments have been increased more higher, HRA should be developed in order to guarantee the better safety of NPP. By the collection of human performance about operators in main control room of NPP, especially in accident situation, it is very important to enhance the human reliability. This paper chooses Loss of Coolant Accident (LOCA) and Steam Generator Tube Rupture (SGTR) as the initiating events, and base on this, some loss of other equipment or system have been added. Then the process that the operators deal with the accidents has been recorded in the accident situation in order to get reasonable human performance data. After we analyzed all of these video by some tools such as eye tracking tool, some items have been found which are very important to the human reliability such as training level, operation task, human-machine interface, surrounding, team work, etc. According to the analysis of the human performance result, we have evaluated the computerized human machine interface. As a result, it also could be benefit to improve the design of NPP, specially the design of main control room.
Results of the LIVE-L4 Experiment on Melt Behavior in the RPV Lower Head
Authors: Miassoedov Alexei , Cron Thomas , Foit Jerzy , Gaus-Liu Xiaoyang , Palagin Alexander , Schmidt-Stiefel Silke , Wenz Thomas
Published: April 2011

The development of a corium pool in the lower head and its behavior is still a critical issue and is of great importance to assess the severe accident progression consequences to ensure the nuclear plant safety. Therefore, experimental efforts are a vital element of the assessment process, providing hard data and insights of the complicated multi-component, highly turbulent corium pool dynamics. It is essential to consider the whole evolution of the accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behavior after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at the Karlsruhe Institute of Technology (KIT) is to study these phenomena experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behavior. The LIVE-L4 experiment was performed using a non-eutectic melt (KNO3-NaNO3) as a simulant fluid. Besides the transient behavior, for which the LIVE-L4 test provides qualified data on temperature evolution in the molten pool and crust growth rates, the experiment addresses other important phenomena, such as the local distribution of heat flux, and the influence of solidification on the thermal-hydraulics of the pool, i.e. the possible existence of a mushy region and its impact on the heat transfer. In the post-test analysis crust thickness profile along the vessel wall, the crust composition and the morphology were determined. The results of this experiment also allow a comparison with findings obtained earlier in other experimental programs. The LIVE-L4 experimental results are being used for the assessment of correlations and development and validation of mechanistic models for the description of molten pool behavior. These calculations are complemented by analyses with the CFD code CONV (thermal hydraulics of heterogeneous, viscous and heat-generating melts) which was developed at IBRAE. The CONV code was applied to simulate the LIVE-L4 test: a) assuming homogeneous heat generation in the liquid and b) accounting for wire heaters used to simulate the heat generation in the melt. Though the results of calculations demonstrate satisfactory agreement with the experimental measurements, deficiencies in the code prediction have been identified regarding e.g. the prediction of the crust thickness. The paper summarizes the objectives of the LIVE program, the main results obtained in the LIVE-L4 experiment and the results of the post-test calculations performed with the CONV code.
Thermal Performance of Vertical Rectangular Channel Inserting Copper Wire With High Porosity
Authors: Takeda Tetsuaki , Honma Akira , Sugai Yuta
Published: April 2011

In the Very-High-Temperature Reactor (VHTR) which is the next generation nuclear reactor system, ceramics and graphite are used as the fuel coating material and the core structural material, respectively. Even if the accident occurs and the reactor power goes up instantly, the temperature of the core will change slowly. This is because the thermal capacity of the core is so large. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel (RPV). From the view point of the safety characteristic, the passive cooling system should be designed for the VHTR as the best way of the reactor and vessel cooling systems (VCS). So, the gas cooling system by natural convection is the one of the candidate systems for the VCS of the VHTR. This study is to develop the passive cooling system for the VHTR using the vertical rectangular channel inserting porous materials. In general, when the high temperature circular or rectangular channels are cooled by forced convection of gas, there are several methods for enhancement of heat transfer such as attaching radial or spiral fins on a channel surface or inserting twisted tape in a channel. The objective of this study is to investigate heat transfer characteristics by forced convection of porous materials inserted into a rectangular channel with high porosity. In order to obtain the heat transfer characteristics of the one-side heated vertical rectangular channel inserting the porous material, an experiment was carried out. From the results obtained in this experiment, it was found that an amount of removed heat by forced convection using porous material (porosity > 0.996) was about 15% higher than that without the copper wire. Furthermore, the ratio between the amounts of heat removed of the rectangular channel with the porous material and without the porous material increases with increasing temperature of the channel wall.
High Performance Reconfigurable Hardware Acceleration on Neutron Transport Computation Based on AGENT Methodology
Authors: Xiao Shanjie , Jevremovic Tatjana
Published: April 2011

A high performance hardware acceleration coprocessor built on field programmable arrays (FPGAs) is designed to accelerate neutron transport computation for three dimensional whole reactor cores. The acceleration coprocessor is designed based on the reconfigurable computation techniques and adopts the dataflow-driven non von Neumann architecture for high efficient parallel computation. The hardware acceleration coprocessor supports much more intensive available computation power compare with the same-era CPUs, and is compatible with existing software acceleration methods. It reaches about 20 times speed up in simulation validations. It is the first time that the reconfigurable hardware acceleration techniques are used to improve the computational efficiency of the reactor physics and neutron transport simulations.
Evaluation on Radiological Consequence Induced by MSLB Accident With Alternative Source Term
Authors: Li Jingxi , Huang Gaofeng , Tong Lili
Published: April 2011

The major threat that nuclear power plants (NPPs) pose to the safety of the public comes from the large amount radioactive material released during design-basis accidents (DBAs). Additionally, many aspects of Control Room Habitability, Environmental Reports, Facility Siting and Operation derive from the design analyses that incorporated the earlier accident source term and radiological consequence of NPPs. Depending on current applications, majority of Chinese NPPs adopt the method of TID-14844, which uses the whole body and thyroid dose criteria. However, alternative Source Term (AST) are commonly used in AP1000 and some LWRs (such as Beaver Valley Power Station, Units No. 1 and No. 2, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 And 2, Kewaunee Power Station and so on), so it is attempted to adopt AST in radiological consequence analysis of other nuclear power plants. By introducing and implementing the method of AST defined in RG 1.183 and using integral safety analysis code, a pressurized water reactor (PWR) of 900 MW nuclear power plant analysis model is constructed and the radiological consequence induced by Main Steam Line Break (MSLB) accident is evaluated. For DBA MSLB, the fractions of core inventory are assumed to be in the gap for various radionuclides and then the release from the fuel gap is assumed to occur instantaneously with the onset of assumed damage. According to the assumptions for evaluating the radiological consequences of PWR MSLB, dose calculation methodology is performed with total effective dose equivalent (TEDE) which is the criteria of dose evaluation. Compared with dose criteria of RG 1.183, the dose of control room, exclusion area boundary and outer boundary of low population zone are acceptable.
Numerical Simulation of 3D Liquid Sloshing Motion With Solid Particles Using Finite Volume Particle Method
Authors: Guo LianCheng , Zhang Shuai , Morita Koji , Fukuda Kenji
Published: April 2011

Sloshing dynamics of a molten core is one of the fundamental behaviors in core disruptive accidents of a liquid-metal cooled reactor. In addition, solid particle-liquid mixture comprising molten fuel, molten structure, refrozen fuel, solid fuel pellets, etc. could lead to damping of its flowing process in a disrupted core. The objective of the present study is to investigate the applicability of the finite volume particle method (FVP), which is one of the moving particle methods, to 3D motion of liquid sloshing processes measured in a series of experiments. In the first part of this study, a typical sloshing experiment of single liquid phase is simulated to verify the present 3D FVP method for sloshing characteristics that include free surface behaviors. Second, simulations of sloshing problems with solid particles are performed to validate the applicability of the FVP method to the 3D motion of solid particle-liquid mixture flows. Some good agreements between the simulation and its corresponding experiment demonstrate applicability of the present FVP method to 3D fluid dynamics of liquid sloshing flow with solid particles.
Experimental Study on Jet Breakup Behavior With Surface Solidification
Authors: Wada Takashi , Abe Yutaka , Kaneko Akiko , Uchiyama Yuta , Nariai Hideki , Yamagishi Makoto , Koyama Kazuya , Itoh Kazuhiro
Published: April 2011

For the safety design of the Fast Breeder Reactor (FBR), the Post Accident Heat Removal (PAHR) is required when a hypothetical Core Disruptive Accident (CDA) occurs. In the PAHR, it is strongly required that the molten core material can be cooled down and solidified by the sodium coolant in the reactor vessel. There is high possibility for molten material to be ejected as a liquid jet into sodium coolant in the reactor vessel. In order to estimate whether the molten material jet is completely solidified by sodium coolant or not, it is necessary to understand the interaction between molten core material and coolant such as jet breakup and fragmentation behavior in coolant. The jet breakup behavior is the phenomenon that the front of molten material breaks up in coolant. To clarify the mechanism of jet breakup and fragmentation during the CDA for the FBR, it is necessary to understand the correlation between jet breakup lengths and size distribution of fragments when molten material jet interacting with coolant. The objective of the present study is to clarify the dominant factor of the jet breakup length and the size distribution of fragments experimentally. Molten jet of U-alloy 138 is injected into water as simulated core material and coolant by free-fall. The density ratio of core material and coolant is almost same as that of the real FBR system. The jet breakup behavior as interaction of molten material with coolant is observed with high speed video camera. Front velocity of the molten material jet is estimated by using the image processing technique. It suddenly decreases when the jet fall into the coolant. The jet breakup length estimated from observed images is compared with the breakup theories to understand the effect of experimental parameters for the jet breakup length. The solidified fragments are gathered and classified in size, and the mass in each size is measured. Median diameter is obtained from the mass distribution of the fragments. In comparison with interfacial instabilities, the median diameter of fragments shows the independent of relative velocity. The jet breakup lengths and median diameters compared with existing theories is discussed.
Status and Proposal Studies on Thermal-Hydraulic Phenomena to Steam Generator Tube Rupture Accident in LFR and ADS
Authors: Sa Rongyuan , Takahashi Minoru
Published: April 2011

This article reviews the studies on steam generator tube rupture (SGTR) accident in lead alloy-cooled fast reactors (LFR) and the accelerator-driven transmutation system (ADS) including proposal for visualization test in injection of water into molten lead-bismuth eutectic (LBE) and lead (Pb) using neutron radiography. The SGTR and the consequent interaction between LBE/Pb and water, i. e., coolant-coolant interactions (CCI), represents an important concern for the safety of LFR and ADS since the CCI might bring a local core voiding, and even initiate a sloshing motion of liquid metal coolant to damage the vessel, which is similar with fuel-coolant interaction (FCI) in light water reactors (LWR) and sodium-cooled fast reactors. Based on the review of related studies, visualization test in special conditions using neutron radiography is proposed to investigate this transient thermal-hydraulic phenomenon.
The Development and Innovation of Spent Fuel Reprocessing in Fuel Cycle
Authors: Chen Guang Jun , Cui Yu Lin , Zhang Guo Guo , Yao Hong Jun
Published: April 2011

With an increased population and an increasing demand for power, nuclear power has attracted an increasing attention and mass nuclear power plant have been built in different countries in the past several decades. At present, about ten thousands ton spent fuels are discharged from nuclear power plant every year and the estimated capacity will approximately add up to 5×105 ton. Therefore, spent fuel reprocessing, by which the co-extraction and separation as well as purification of Uranium and Plutonium could be realized and ensure the recycle of uranium resources and the management of nuclear waste, is a vital step in nuclear fuel cycle including two major strategies, i.e. once-through cycle and closed fuel cycle. It is worth noting that the utilization of MOX fuel made by plutonium mixed with uranium has been successfully achieved in thermal reactor. Fortunately, the middle experiment plant of china spent fuel reprocessing has been being debugged and will be operated completely in future two years. Various reprocessing schemes have been proposed for the extraction of actinides from fission products and other elements presented in spent nuclear fuel. However, after numerous studies of alternate reprocessing methods and intensive searches for better solvents, the PUREX process remains the prime reprocessing method for spent nuclear fuels throughout the world. High burning and strong radioactive spent fuel resulting from the evolution of various reactors drive the development of the advanced PUREX technology, which emphasizes the separation of neptunium and technetium besides the separation of the Uranium and Plutonium from the majority of highly active fission products. In addition, through Partitioning and Transmutation method, some benefits such as segregating the actinides and long life fission products from the high level waste can be obtained. The GANEX process exploited by CEA, which roots in COEX process belonged to advanced PUREX process, considers the separation of the actinides and long life fission products. The study on the pyro-chemical processing such as the method of electro-deposition from molten salts has still not replaced the traditional PUREX process due to various reasons. In conclusion, the future PUREX process will focus on the modified process including predigesting the technical flowsheets and reducing reprocessing costs and using salt-less reagent in order to minimize the waste production.
Measurement of Wavy Surface Oscillations on Liquid Metal Lithium Jet for IFMIF Target
Authors: Sugiura Hirokazu , Kanemura Takuji , Yoshihashi-Suzuki Sachiko , Kondo Hiroo , Yoshikawa Tomohide , Yamaoka Nobuo , Ida Mizuho , Nakamura Hiroo , Matsushita Izuru , Horiike Hiroshi
Published: April 2011

This paper reports on the measurement of surface waves on a liquid lithium jet and results of the study of the Li target at the International Fusion Materials Irradiation Facility (IFMIF). The characteristics of the surface waves at the nozzle exit and just downstream of it were examined experimentally, since the initial growth of free surface waves exerts a definite influence on surface behavior in the downstream region. Experiments were carried out using the lithium circulation loop at Osaka University, with a focus on the free surface oscillations. These oscillations were measured using an electrocontact probe apparatus, which detects electric contacts between the probe tip and the Li surface. The apparatus was installed 15 mm downstream from the nozzle exit and was scanned along the liquid-depth direction. The contact signal recorded in the experiment was analyzed, and the wave amplitude and frequency of the surface waves were examined.
Effect of Gravity on Phase Distribution Patterns of Bubbly Two-Phase Flow in a Vertical Mini Pipe
Authors: Watanabe Shunichi , Fukuhara Yutaka , Hazuku Tatsuya , Takamasa Tomoji , Hibiki Takashi
Published: April 2011

In relation to the development of the interfacial area transport equation, a precise database of the axial development of void fraction profile, interfacial area concentration and Sauter mean bubble diameter in an adiabatic nitrogen-water bubbly flow in a 5 mm-diameter mini pipe was constructed for normal and microgravity conditions using stereo image-processing. The flow measurements were performed at four axial locations (axial distance from the inlet normalized by the pipe diameter, z/D = 5.5, 34, 72 and 110) and with various flows: superficial gas velocity of 0.00434–0.0420 m/s, and superficial liquid velocity of 0.239–0.949 m/s. The effect of gravity on radial distribution of bubbles and the axial development of two-phase flow parameters is discussed in detail based on the obtained database and visual observation.
Analysis of Fuel Coolant Interaction in PWR With Experiment and Computer Code
Authors: Wang Xi , Li Tian-shu , Huang Xi , Yang Yan-hua
Published: April 2011

Fuel-Coolant Interaction can be divided into two stages including premixing and explosion. In the stage of premixing, the liquid corium flowing from the core fragmented into drops, and the drops may fragmented into fine fragmentations leads to steam explosion in the condition of triggering. In the severe accident, the contact of melt-fuel and coolant may result into steam explosion, which may threaten integrity of containment. The prediction of steam explosion must be based on results of experimental research and on simulations with computer code. In this work, the experimental result of low temperature fuel-coolant interaction is compared with the computing result of FCI-CMFD code MC3D, analyzing the phenomenon of FCI, the diameter of droplet and the pressure wave of explosion. Then, the MC3D is used in the analysis and prediction of fuel coolant interaction in CPR1000 NPP model.
Development of System Based Code: Part 2—Application of Reliability Target for Configuration of ISI Request
Authors: Takaya Shigeru , Okajima Satoshi , Kurisaka Kenich , Asayama Tai , Machida Hideo , Kamishima Yoshio
Published: April 2011

The System Based Code (SBC) concept has been proposed to achieve compatibility in matters of reliability, safety, and cost of Fast Breeder Reactors (FBR). This code extends the present structural design standard to include the areas of load setting, fabrication, inspection, maintenance, and so on. Therefore, a quantitative index which can connect different areas is required. In addition, the determination of its target value is also one of the key points to substantiate the SBC concept. Failure probability is one of the candidate indexes. We have proposed a new method to determine the reliability targets for the structures and components in FBR plants from the safety point of view by utilizing analysis models of a probabilistic safety assessment. In this study, the effectiveness of the failure probability as an index and the compatibility of the reliability targets derived by the new method were investigated through a trial setting of In-Service Inspection (ISI) request on the reactor vessel near the sodium surface based on the SBC concept. The failure probability due to creep-fatigue interaction was calculated by the Monte-Carlo simulation. In response, the reliability targets for fracture related to the risk from internal initiating events were derived. Cumulating the failure probability and the reliability targets up to the end of in-service period enables us to compare them directly, and we obtained a result that the reactor vessel has enough reliability even without ISI. Through this trial, we showed that the failure probability is a promising index, and the reliability targets derived by the new method are compatible with the SBC concept.
Numerical Simulation of Cavitation for Comparison of Sodium and Water Flows
Authors: Ardiansyah Teddy , Takahashi Minoru , Yoshizawa Yoshio , Nakagawa Masamichi , Asaba Makoto , Miura Kuniaki
Published: April 2011

In this paper, numerical simulation of cavitation for water and sodium flows were performed by using CFD code package FLUENT and the results were compared with the experiments. The geometry of the grid for numerical calculation was a venturi with ID and OD of 6.5 and 21 mm, respectively. The numerical simulations showed that the onset cavitation conditions were affected significantly by the non-condensable gas content in the liquid. Comparing with the experimental results, the non-condensable gas contents that gave close predictions of onset cavitation conditions with the experiments were larger in water than in liquid sodium because of the different solubility property of air and argon gas. Parametrical analysis showed that the void fraction distributions for water and liquid sodium did not show large differences in case of the same caviation coefficient and non-condensable gas contents.
Achieving Resource Sustainability in China Through the Thorium Fuel Cycle in the Candu Reactor
Authors: Boczar Peter G. , Hyland Bronwyn , Bradley Keith , Kuran Sermet
Published: April 2011

The CANDU® reactor is the most resource-efficient reactor commercially available. The features that enable the CANDU reactor to utilize natural uranium facilitate the use of a wide variety of thorium fuel cycles. In the short term, the initial fissile material would be provided in a “mixed bundle”, in which low-enriched uranium (LEU) would comprise the outer two rings of a CANFLEX® bundle, with ThO2 in the central 8 elements. This cycle is economical, both in terms of fuel utilization and fuel cycle costs. The medium term strategy would be defined by the availability of plutonium and recovered uranium from reprocessed used LWR fuel. The plutonium could be used in Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. Recovered uranium could also be effectively utilized in CANDU reactors. In the long term, the full energy potential from thorium could be realized through the recycle of the U-233 (and thorium) in the used CANDU fuel. Plutonium would only be required to top up the fissile content to achieve the desired burnup. Further improvements to the CANDU neutron economy could make possible a very close approach to the Self-Sufficient Equilibrium Thorium (SSET) cycle with a conversion ratio of unity, which would be completely self-sufficient in fissile material (recycled U-233).
Numerical Simulation of Vortex Shedding From Elongated Rectangular Cylinders in a Rectangular Channel
Authors: Liu Lifang , Lu Daogang , Li Quanxing
Published: April 2011

A two-dimensional code was developed to simulate vortex shedding characteristic and flow-structure interaction (FSI) of plate-type structures. In the code the physical component boundary fitted coordinate (PCBFC) was used to deal with the curve boundary. The arbitrary Lagrangian Eulerian (ALE) method was used to realize the grid movement. A barrier unit idea was adopted to deal with the boundary of fluid domain and solid domain in the code. The code was validated by comparing the numerical simulation results with experimental data. It was found that the vortex shedding phenomena in case of rectangular cylinder are strongly related to the length of the rectangular cylinder in the stream line.
Flow Mixing Inside a Control-Rod Guide Tube: Part II—Experimental Tests and CFD-Simulations
Authors: Angele Kristian , Cehlin Mathias , Ho¨gstro¨m Carl-Maikel , Odemark Ylva , Henriksson Mats , Tinoco Hernan , Lindqvist Hans , Hemstro¨m Bengt
Published: April 2011

A large number of control rod cracks were detected during the refuelling outage of the twin reactors Oskarshamn 3 and Forsmark 3 in the fall of 2008. The extensive damage investigation finally lead to the restart of both reactors at the end of 2008 under the condition that further studies would be conducted in order to clarify all remaining matters. Also, all control rods were inserted 14% in order to locate the welding region of the control rod stem away from the thermal mixing region of the flow. Unfortunately, this measure led to new cracks a few months later due to a combination of surface finish of the new stems and the changed flow conditions after the partial insertion of the control rods. The experimental evidence reported here shows an increase in the extension of the mixing region and in the intensity of the thermal fluctuations. As a part of the complementary work associated with the restart of the reactors, and to verify the CFD simulations, experimental work of the flow in the annular region formed by the guide tube and control rod stem was carried out. Two full-scale setups were developed, one in a Plexiglass model at atmospheric conditions (in order to be able to visualize the mixing process) and one in a steel model to allow for a higher temperature difference and heating of the control rod guide tube. The experimental results corroborate the general information obtained through CFD simulations, namely that the mixing region between the cold crud-removal flow and warm by-pass flow is perturbed by flow structures coming from above. The process is characterized by low frequent, high amplitude temperature fluctuations. The process is basically hydrodynamic, caused by the downward transport of flow structures originated at the upper bypass inlets. The damping thermal effects through buoyancy is of secondary importance, as also the scaling analysis shows, however a slight damping of the temperature fluctuations can be seen due to natural convection due to a pre-heating of the cold crud-removal flow. The comparison between numerical and experimental results shows a rather good agreement, indicating that experiments with plant conditions are not necessary since, through the existing scaling laws and CFD-calculations, the obtained results may be extrapolated to plant conditions. The problem of conjugate heat transfer has not yet been addressed experimentally since complex and difficult measurements of the heat transfer have to be carried out. This type of measurements constitutes one of the main challenges to be dealt with in the future work.
A Simplified Model for a Preliminary Study of the Dynamic Behaviour of a Small Gen IV LFR DEMO
Authors: Bortot Sara , Cammi Antonio , Artioli Carlo , Camprini Patrizio Console , Ghazy Rasha
Published: April 2011

The Lead Cooled Fast Reactor (LFR) is one of the six concepts selected by the Generation IV International Forum (GIF) as candidates for the long term evolution of nuclear technology. Due to the significant technological innovations it implies, the European Sustainable Nuclear Energy Technology Platform (SNETP) recognized that LFR complete development requires the realization of a demonstration plant (DEMO) as a fundamental intermediate step. In this paper, a preliminary approach to the simulation of DEMO primary system dynamic behavior is presented. The need of investigating reactor responses to temperature transients has led to a simplified model reckoning with all the main feedbacks following a reactivity change in the core, which have been calculated by means of ERANOS deterministic code ver. 2.1 coupled with JEFF3.1 data library. A lumped-parameter approach has been adopted to treat both neutronics and thermal-hydraulics: indeed, the point-kinetics approximation has been employed and an average-temperature heat-exchange model has been implemented. Due to the latter, the dynamic mechanical behavior of DEMO core — modeled as a cylinder — has been addressed by considering expansions and contractions instantaneous with temperature variations, i.e. neglecting mass inertia effects. The very simple linearized model treated in the present work turns out to be a helpful tool in this early phase of the DEMO pre-design, in which all the system specifications are still considered to be open design parameters, since it allows a relatively quick, qualitative analysis of dynamics and stability aspects that cannot be left aside when refining or even finalizing the system configuration.
Maintenance Strategy for Branch Connection on the Hot Leg of Steam Generator in CANDU 6
Authors: Zhao Liang , Hu Jianqun , Wu Zhigang , Wang Xin
Published: April 2011

A maintenance strategy to repair or modify the dissimilar weld on the branch connection of the steam generator hot leg in CANDU6 was developed. The repair method is to add a carbon steel pipe spool as a transition fitting to transfer the dissimilar weld to shop welding, and the modification method is to use the backup tap fabricated during plant installation, so that the weld quality would be superior comparing with field welding, and the quantity of argon gas getting into the primary heat transport system would be reduced simultaneously. A gasbag is used to isolate the branch connection for welding performance when the primary heat transport system is drained to low level state during the prolonged shutdown. If the gasbag could not work, a nonstandard pipe fitting with O-ring should be used to isolate the branch connection.
Improvement of Flange Tightening Method for Small Bore Low Rating Bolted Flanges
Authors: Mori Masako , Nebu Akira , Kanno Takashi
Published: April 2011

An appropriate flange tightening methods for small bore and low rating piping flange joints are clarified to improve the sealing performance of the bolted flange joints and the workability of flange tightening work. It is said that lubricant on the screw of the bolts and the nut-seating surface can minimize the variability of axial force acting on flange bolts, while this process might make it harder to tight the bolts uniformly especially for small bore low rating flanges. So, in this paper the appropriate condition to apply lubricant is clarified by a series of bolt tightening tests and sealing tests results. On the other hand, for the bolted flanges applying spiral wound gaskets, measuring the gasket compress dimensions help us to prevent uneven tightening balance and to perform the appropriate tightening work. Appropriate gasket compress dimensions are also clarified to ensure the sealing performance for the flanges based on the sealing tests results. Based on these test results, recommended flange-tightening methods have been summarized as an instruction and tightening work procedure to improve the sealing performance of the bolted flanges and the workability of flange tightening work.
Subchannel Analysis of Wire Wrapped SCWR Assembly
Authors: Shan Jianqiang , Chen Xuanxiang
Published: April 2011

Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assembly with wire wrap can achieve a more uniform coolant temperature profile than the grid spaced assembly, which will result in a lower peak cladding temperature; (2) The pressure drop in a wire wrapped assembly is less than that in a grid spaced assembly, which can reduce the operating power of pump. (3) The wire wrap pitch has significant effect on the flow in the assembly. Smaller Hwire/Drod will result in stronger cross flow and a more uniform coolant temperature profile, and also a higher pressure drop.
On Line Monitoring the Reactor Power Distribution Based on the Data Library of the Eigenfunctions
Authors: Wang Changhui , Wu Hongchun , Cao Liangzhi
Published: April 2011

A method of on-line monitoring for commercial PWRs using eigenfunctions has been proposed in prevenient works. In this method, the eigenfunctions combine with the detector readings are used to reconstruct the real reactor power distribution. But it is very difficult to choose the eigenfunctions because the condition of the reactor is much complex. Therefore, some improvements on this method are studied in this paper. A number of representational conditions according to the reactor fuel management are picked up to create a data library of different eigenfunctions. On the monitoring process, the computer will judge that which represen6tational conditions most close to the real reactor condition and choose the eigenfunctions which will be used to reconstruct the reactor core power distribution combine with the detector readings. A reactor of Qinshan Nuclear Power Corporation is studied here as an example. The detector readings are from the simulator. The numerical result shows that this method can reconstruct the reactor power distribution with high speed (about 0.03 seconds for each step) and high accuracy (the relative errors are lower than 3% mostly).
Chemico-Physical Causes of Radiation-Induced “Long Cell” Action Corrosion in Water Cooled Reactors
Authors: Saji Genn
Published: April 2011

In the previous overview papers [1, 2], the author has identified that ‘long cell action’ corrosion plays a pivotal role in practically all unresolved corrosion issues, irrespective of reactor types and operation. In trying to confirm the existence of radiation-induced ‘long-cell’ action (macro) corrosion cell in the primary cooling system of LWRs, the author attempted to theoretically reproduce the electrochemical potential difference demonstrated during experiments at the INCA Loop in Sweden and the NRI-Rez Loop in the Czech Republic [3, 4]. By performing a radiation chemistry kinetics study combined with electrochemistry calculations, the hydrated electrons, e−aq, reacting mainly with stable molecules, are found to be responsible for inducing a large portion of the potential difference both in the PWR and BWR water chemistry environment. Considering large uncertainties, the author used the standard equilibrium potential as a fitting parameter in the previous studies [3, 4]. The standard chemical potential of the hydrated electron estimated from the fitting parameter is far less than the generally accepted value of 2.86 V. In order to resolve the large discrepancy between the generally accepted values and the estimation from the fitting parameter, the author has developed a ‘mixed’ radiation-electrochemistry formalism, which enables theoretical reconstruction of the observed potential differences more clearly. The previous verifications are updated by using this approach. Through these studies, the author has confirmed the existence of the ‘long cell’ action corrosion mechanism existing in the water-cooled reactors.
An Experience Study for Advanced MCR of Lungmen Project With Human Factors Regulations: NUREG-0711
Authors: Yenn Tzu-Chung , Chuang Chu-Yu , Hsu Chong-Cheng , Cheng Tsung-Chieh , Chen Ming-Huei , Liou Jin-Liang
Published: April 2011

Nowadays, there is a general consensus that establishing nuclear regulations concerning human factors engineering (HFE) is an important issue. NUREG-0711, original version published in 1994, was developed under such assumptions. And it soon became a common reference for nuclear power plant reviewers and designers. Lungmen NPP is the first ABWR plant in Taiwan and is under construction now. Taipower Company signed the contract for Lungmen Project with General Electric Company in 1995. By Lungmen Project Bid Specification, GE should take the responsibility to design the main control room according to the last version of HFE regulation that is NUREG-0711 version zero. Up to the present, NRC has modified NUREG-0711 twice on the basis of evaluating experiences and users’ feedback from different fields. But the Lungmen NPP has not finished yet. No doubt, the modifications not only make the regulation state-of-the-art but practicable. How to cope with this asynchronous problem between contracts and modification is a critical concern. In this article, we present our resolutions on this issue. Step one; comparing the differences between NUREG-0711 version zero and two. Step two; figuring out what meanings and intent are behind these changes. Step three; following the version zero regulation and taking advanced principle into consideration at the same time. Implementation according to old version regulation and taking the advanced intent and principle from step 2 is a practice resolution from the experience of Lungmen NPP. Those experiences will be helpful for human factors engineering activities on update the advanced main control room of nuclear power plant in the near future.
Sensitivity Analyses of Finite Element Method for Computing Residual Stress of Dissimilar Metal Multi-Pass Weld
Authors: Tsai YaoLong , Li ChaoJen , Tsai TaiPing , Wang Li-Hua , Fan TeWei , Chou ChangPin
Published: April 2011

Recently the stress corrosion cracking cases of dissimilar metal welds are found in many nuclear power plants. Weld-induced residual stress is main factor for crack growth. Therefore exact estimation of residual stress is important for reliable operation. This paper presents residual stress computation performed by EU mock-up Nozzle. Based on ANSYS software, uncoupled thermal-mechanical three-dimensional (3D) and two-dimensional ax symmetric finite element models are developed. Based on 2 dimensional and 3 dimensional finite element analyses, effect of the welding sequence approaches for multi-pass weld such as lump, layer and ring etc. on residual stress variation is estimated for sensitivity analysis.
Evaluation of Safety Culture Promoting Activities: Chubu’s Challenge
Authors: Kurata Satoshi , Suzuki Tetsuya
Published: April 2011

Nuclear safety culture has been one of the most important aspects for safe and reliable operations of nuclear power plants worldwide. Japanese nuclear operators started promoting Safety Culture using the PDCA (Plan-Do-Check-Action) cycle in December 2007. Chubu Electric Power Company has been developing the system of promoting safety culture and challenging to develop evaluation tools for safety culture.
Effect of Melting Processing on Tensile Properties and Microstructure of New RAFM Steel
Authors: Yao Jun , Suo Jinping
Published: April 2011

All of the RAFM steels only safely used under 550°C, that is not enough for the next reactor. An new RAFM steel was melted by non-vacuum induction melting (VIM) and electro-slag remelting (ESR), followed by hot-forging and rolling into rods and plates. In this paper, we investigated the effect of thermal ageing treatment on tensile properties of the rods and plates. The microstructure was studied by OM (optics micrograph) and scanning electron microscopy (SEM). The results showed that by using the same heat treatment process, the tensile strength of the samples was 680MPa, the total elongation was 31%, which were better than the CLAM steel whose tensile strength and total elongation were 668MPa and 25% respectively. The difference between the transverse and the longitudinal properties was reduced markedly. So the ESR played an important part in improving the mechanical properties.
BWR Recirculation Pump Adjustable Speed Drive Performance and Control Affects
Authors: Morgan James W. , Gode Timothy
Published: April 2011

Exelon Nuclear has recently installed adjustable speed drives (ASD), or variable frequency drives (VFD), in place of the original motor-generator (MG) sets for their boiling water reactor (BWR) recirculation pumps at Quad Cities. This paper reviews expected versus actual performance of the drives and motors. The discussion focuses on energy savings, motor starting characteristics, control accuracy and stability, motor and cable thermal behavior; as well as, a comparison of actual supply system input harmonic measurements versus analysis results. Included in the review are a few operational lessons learned with regard to the startup process, input medium voltage low speed holds, loss of cooling water pump suction pressure, cooling system surge tank level indication, and other miscellaneous points of interest.