bluuebunny/crossref_metadata_2025 huggingface dataset:
The Lead-Alloy Corrosion Study at KAERI
Authors: Song Tae Yung , Cho Choonho , Cho Chungho
Published: September 2008

KAERI (Korea Atomic Energy Research Institute) is developing an accelerator-driven transmutation system called HYPER (HYbrid Power Extraction Reactor). HYPER is the 1000MWth system designed to transmute the long-lived TRU (Transuranic Elements) and FP (Fission Product) included in the PWR spent fuel. LBE (Lead-Bismuth Eutectic) is used as the spallation target and coolant material in HYPER. KAERI has also investigated the conceptual design of a lead-cooled fast reactor. Lead (Pb) is used as the coolant material in that reactor. The most significant problem is a corrosion when using the lead-alloy liquid metal in those reactors. Therefore, it is necessary to study the corrosion characteristics and develop the technology to protect the steel structure materials against a corrosion. KAERI has been developing the facilities needed to study the corrosion of lead-alloy. KAERI fabricated a static corrosion test facility in 2003. The static corrosion tests of HT-9, 316L and T91 have been performed at 600 °C and 650 °C since 2003. The Pb-Bi loop was constructed in 2006. The Pb-Bi loop is an isothermal loop which can be operated at temperatures up to 550 °C. The Pb loop is designed to be operated with ΔT = 150 °C (Tmin = 450 °C and Tmax = 600 °C). The first stage of the Pb loop construction was finished and operations began in 2006. We will complete the second stage of the Pb loop construction after testing the first stage Pb loop.
Put a Coalatom in Your Tank: The Compelling Case for a Marriage of Coal and Nuclear Energy
Authors: Penfield Scott R. , Bolthrunis Charles O.
Published: September 2008

Increasing costs and security concerns with present fossil energy sources, plus environmental concerns related to CO2 emissions and the emergence of new technologies in the energy and transportation sectors set the stage for a marriage of convenience between coal and nuclear energy. As the price of oil continues to increase and supply becomes increasingly constrained, coal offers a secure domestic alternative to foreign oil as a source of liquid fuels. However, conventional technologies for converting coal to liquid fuels produce large quantities of CO2 that must be released or sequestered. Advanced nuclear technologies, particularly the High-Temperature Gas-Cooled Reactor (HTGR), have the potential to produce hydrogen via water splitting; however, the transportation and storage of hydrogen are significant barriers to the “Holy Grail”, the Hydrogen Economy. In a coal/nuclear marriage, the hydrogen and oxygen provided by nuclear energy are joined with coal as a source of carbon to provide liquid fuels with negligible CO2 release from the process. In combination with emerging hybrid vehicles, fuels based on a coal/nuclear marriage promise stable prices, increased domestic security and a reduction in CO2 emissions without the need to completely replace our transportation fuels infrastructure. The intent of this paper is to outline the technical basis for the above points and to show that process energy applications of nuclear energy can provide the basis for answering some of the tougher questions related to energy and the environment.
Sodium-Water Reaction and Thermal Hydraulics at Gas-Liquid Interface: Numerical Interpretation of Experimental Observations
Authors: Yamaguchi Akira , Takata Takashi , Ohshima Hiroyuki , Suda Kazunori
Published: September 2008

In a sodium-cooled fast reactor development, coupled phenomena of thermal-hydraulics and chemical reaction of sodium and water vapor are of importance from the safety viewpoint. However, the sodium-water reaction (SWR) phenomena are generally complex and the experimental measurement technology is not well matured. Therefore, a numerical simulation is used for the investigation of the SWR. In this paper, a new computer program has been developed and the SWR in a counter-flow diffusion flame is studied by a numerical simulations and an experiment as well. In the computer program, Navier-Stokes equations and chemical reaction equations are solved interactively. In addition, a dynamic equation of airborne particulates is coupled with the governing equations of thermal hydraulics. A source of the particulates is the chemical reaction products, i.e. sodium hydroxide and sodium oxide. The SWR experiment is decided based on the numerical simulation. To obtain a stable reaction flame and to measure the temperature and reaction product distributions, the flow field in the experimental cell needs to be optimized. The numerical simulation is useful for designing experiments of complex phenomena and for obtaining the data. The computations are compared with experimental data. It has been demonstrated that the computational fluid dynamics code coupled with chemical reaction well predict the SWR.
A Methodology for Discerning Incipient Boiling of the Coolant in a Water-Moderated, Water-Cooled (Pressurized-Water) Nuclear Reactor by Means of the Bayesian Neutron-Noise Classifier
Authors: Sharayevskij Igor
Published: September 2008

Noise signals at the output of the technological parameters of a nuclear power plant (neutron flux, dynamic pressure, etc.) contain important information on the stale of the equipment. Effective algorithms have been developed for identifying random processes which, after appropriate transformation, are treated as multidimensional random vectors. These vectors are automatically classified by means of the aforementioned algorithms on the basis of the likelihood function, in particular, of the Bayesian classifier and resolving functions. The use of the Bayesian classifier is based on constructing multidimensional distributions of probabilities in the attribute space with the applicable dimensionality, which serves to describe random representing vectors of the applicable images that are subject to automatic classification.
A Review of the Effects of Elevated Temperature on Concrete Materials and Structures
Authors: Naus D. J. , Graves H. L.
Published: September 2008

Concrete’s properties are more complex than those of most materials because not only is concrete a composite material whose constituents have different properties, but its properties depend upon moisture and porosity. Exposure of concrete to elevated temperature affects its mechanical and physical properties. Elements could distort and displace, and, under certain conditions, the concrete surfaces could spall due to the buildup of steam pressure. Because thermally-induced dimensional changes, loss of structural integrity, and release of moisture and gases resulting from the migration of free water could adversely affect plant operations and safety, a complete understanding of the behavior of concrete under long-term elevated-temperature exposure as well as both during and after a thermal excursion resulting from a postulated design-basis accident condition is essential for reliable design evaluations and assessments of nuclear power plant structures. As the properties of concrete change with respect to time and the environment to which it is exposed, an assessment of the effects of concrete aging is also important in performing safety evaluations. The effects of elevated temperature on Portland cement concretes and constituent materials are summarized, design codes and standards identified, and considerations for elevated temperature service noted.
The Fate of Hazardous Materials in the Biosphere
Authors: Sandquist Gary M.
Published: September 2008

Although nuclear power appears to be expanding as a major global energy source, the disposal of radioactive waste from the nuclear fuel cycle still poses formidable challenges to the full expansion of the nuclear enterprise. The perception that nuclear wastes represent unique and insoluble threats to humans is ill founded. The risk from these radioactive materials is comparable and many ways less severe than other more familiar hazardous materials that are ubiquitous in the biosphere. Radioactive materials decay and reduce in time unlike stable elements. Besides the reduction of radioactive materials through decay, the dilution and dispersion of all hazardous materials by natural forces and events provides the reduction required to make adequate and safe disposal of nuclear waste possible. The ultimate sink for essentially all of these hazardous wastes will prove to be the oceans with their great capacity of dilution and containment.
Measurement of the Velocity and Pressure Drop in a Tubular Type Fuel
Authors: Park Jonghark , Chae Heetaek , Park Cheol , Kim Heonil
Published: September 2008

We have developed a tubular type fuel assembly design as one of candidates for fuel to be used in the Advanced HANARO Reactor (AHR). The tubular type fuel has several merits over a rod type fuel with respect to the thermal-hydraulic and structural safety; the larger ratio of surface area to volume makes the surface temperature of a fuel element become lower, and curved plate is stronger against longitudinal bending and vibration. In the other side, a disadvantage is expected such that the flow velocity can be distributed unevenly channel by channel because the flow channels are isolated from each other in a tubular type fuel assembly. In addition to the design development, we also investigated the flow characteristics of the tubular fuel experimentally. To examine the flow velocity distribution and pressure drop, we made an experiment facility and a mockup of the tubular fuel assembly. The fuel assembly consists of 6 concentric fuel tubes so that 7 layers are made between fuel tubes. Since each layer is divided into three sections by stiffeners, 21 isolated flow channels are made in total. We employed pitot-tubes to measure the coolant velocity in each channel. The maximum velocity was measured as large as about 28% of the average velocity. It was observed in the innermost channel contrarily to the expectation from the hydraulic diameter. A change in the total flow rate did not affect the flow distribution. Meanwhile, the pressure drop was measured as about 70% of the drop in the rod type fuel assembly in use in HANARO.
Factors to Be Considered for Selection and Design of Rupture Disc
Authors: Murty D. V. R. , Reddy A. Chennakeshava , Bhaskar Rao Y.
Published: September 2008

All over the world numerous fatalities and large property losses have taken place because of Industrial Explosions. In order to mitigate the losses of life and property due to overpressure rupture discs are used. Successful implementation would depend on many factors and one of the important factors is the selection and design of rupture discs. There are several factors that affect the disc performance. The present paper discusses the factors that effect the disc performance so that the right type of disc can be selected.
The TACIS Nuclear Programme: Assistance in Upgrading Russian Nuclear Power Stations —An Overview of the Individual Projects in the Internet
Authors: Bie`th Michel , Schoels Hubert
Published: September 2008

The European Union’ TACIS programme has been established for the New Independent States (NIS), among them in the Russian Federation since 1991. One priority of TACIS funding is Nuclear Safety. The European Commission has made available a total of 944 Million € for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. JRC is carrying out works in the following areas: • On-Site Assistance for TACIS operating Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER; • Regulatory Assistance; • Industrial Waste Management; • Nuclear Safeguards. All TACIS projects, dealing with these areas of activity are now available in so called Project Description Sheets (PDS) or Project Results Sheets (PRS) in the Internet for everybody. JRC has created in the Internet an easy to open and to browse database which contains the result of works in relation to the above mentioned nuclear activities. This presentation gives an on-line overview of the app. 430 projects which have been implemented so far since the outset of the TACIS Nuclear Progremme in the Russian Federation, which is representative to the other CIS countries, benefiting from the TACIS. The presentation will mainly consist of an on-line-demonstration of the TACIS Nuclear WEB Page, created by JRC.
Modal Wavelets Analysis to Gas-Liquid Two Phase Flow PIV Images
Authors: Takei Masahiro , Hassan Yassin A. , Ortiz-Villafuerte J. , Uemura Tomomasa
Published: September 2008

A modal wavelet transform, which overcomes the intrinsic data number limitation of power of two to conventional wavelet transform, has been applied to analysis of pseudo and real bubbly flow PIV images. The modal wavelet transform is compared with the discrete wavelet transform in order to select the best base function among Neumann, Dirichlet and Green function types base functions. Consequently, it is verified that Neumann type base function is the best because the correlation of Neumann type base function is the highest. From the result of wavelet analysis of the real bubbly flow PIV image, as the relative velocity is higher, the dominant eddy scale becomes smaller. The extraction modal wavelet level depends on the base function.
Technology Efficiency Study on Nuclear Power and Coal Power in Guangdong Province Based on DEA
Authors: Li Yinong , Wang Dong
Published: September 2008

Guangdong Province has taken the lead in embarking on nuclear power development to resolve its dire lack of primary resources. With the deepening of the on-going structural reform in the electric power sector in China, the market competition scheme is putting electricity generation enterprises under severe strain. Consequently, it is incumbent upon the nuclear power producers to steadily upgrade management, enhance technical capabilities, reduce cost and improve efficiency. At present, gradual application of such efficiency evaluation methodology has already commenced in some sectors in China including the electric power industry. The purpose of this paper is to use the Data Envelopment Analysis (DEA), which is a cutting-edge approach in the efficiency evaluation field — to study the technological efficiency between nuclear power and coal power in Guangdong Province. The DEA results demonstrate that, as far as Guangdong Province is concerned, the technological efficiency of nuclear power is higher than that of coal power in terms of Technological Efficiency (TE), Pure Technology Efficiency (PTE) and Scale Efficiency (SE). The reason is that nuclear power technology is advanced with a much higher equipment availability factor. Under the same scale, the generation output of nuclear power is far higher than that of equivalent coal power generation. With the envrionemntal protection and sustainable development requirements taken into full account, nuclear power constitutes a clean, safe and highly-efficient energy form which should be extensively harnessed in Guangdong Province to fuel its future continuing economic growth.
Mitigation of Anticipated Transients Without Scram in the ESBWR
Authors: Marquino Wayne , Shiralkar Bharat , Sitaraman Shivakumar
Published: September 2008

The ESBWR has robust ATWS performance due to the use of passive systems. The heatup of the suppression pool is minimized and the ATWS can be managed at high pressure. ATWS prevention/mitigation features of ESBWR include: 1) An Alternate Rod Insertion (ARI) system that utilizes sensors, logic and valves that are diverse and independent of the Reactor Protection System RPS, 2) Electrical insertion of Fine Motion Control Rod Drives (FMCRDs) that utilize sensors and logic which are diverse and independent of the RPS, 3) Automatic feedwater runback under conditions indicative of an ATWS, 4) An automatic Standby Liquid Control System (SLCS) with a capacity equivalent to 5.42E−3 m3/sec (86 gpm) of 13-weight percent sodium pentaborate solution. The boron is supplied by 2 accumulators and injection is triggered by Squib valves. The boron solution is piped inside the reactor to the core bypass, and discharges at several distribution points in high velocity jets. Several other ESBWR features help to mitigate an ATWS. The large steam volume in the chimney results in a reduced vessel pressurization rate. The isolation condensers terminate suppression pool heatup while there is still steam generation occurring in the core. Analysis with the TRACG computer code shows the ESBWR safety valve capacity is effective in limiting the pressure in the reactor vessel pressure. The automatic FW reduction reduces water level and core flow, which in turn reduces power to approximately one-quarter of the initial value. Finally the reactor is brought sub-critical by boron injection into the core bypass region from two Standby Liquid Control System accumulators. The results show that the mitigation systems are adequate to meet the acceptance criteria without operator action.
On the Accuracy Evaluation of Ultrasonic Doppler Flowmeter
Authors: Ohkubo Tomoyuki , Takeda Yasushi , Mori Michitsugu
Published: September 2008

The accuracy evaluation of a pipe flowmeter using ultrasonic velocity profiler is investigated theoretically and experimentally. The error depends basically on the data points on discretized velocity profile, but it decreases rapidly with increasing the data points. It can be reduced, relative to the theoretical velocity profile, below 1% with about 100 data points. The error arising from averaging the instantaneous flow rate increases with the fluctuation amplitude in the loop and decreases with its frequency and total measurement time. A procedure to determine the total averaging time is proposed.
Modelling of Irradiated Materials
Authors: Dutta B. K. , Durgaprasad P. V. , Pawar A. K. , Kushwaha H. S. , Banerjee S.
Published: September 2008

Irradiation of materials by energetic particles causes significant degradation of the mechanical properties, most notably an increased yield stress and decrease ductility, thus limiting lifetime of materials used in nuclear reactors. The microstructure of irradiated materials evolves over a wide range of length and time scales, making radiation damage and inherently multi-scale phenomenon. At atomic length scale, the principal sources of radiation damage are the primary knock-on atoms that recoil under collision from energetic particles such as neutrons or ions. These knock-on atoms in turn produce vacancies and self-interstitial atoms, and stacking fault tetrahedra. At higher length scale, these defect clusters form loops around existing dislocations, leading to their decoration and immobilization, which ultimately leads to radiation hardening in most of the materials. All these defects finally effect the macroscopic mechanical and other properties. An attempt is made to understand these phenomena using molecular dynamics studies and discrete dislocation dynamics modelling.
Nuclear-Driven Copper-Based Hybrid Thermo/Electro Chemical Cycle for Hydrogen Production
Authors: Khalil Yehia F. , Rostkowski Katherine H.
Published: September 2008

With a worldwide need for reduction of greenhouse gas emissions, hydrogen gas has become a primary focus of energy researchers as a promising substitute of nonrenewable energy sources. For instance, use of hydrogen gas in fuel cells has received special technological interest particularly from the transportation sector, which is presently dominated by fuel oil. It is not only gaseous hydrogen that is in demand, but the need for liquid hydrogen is growing as well. For example, the aerospace industry uses liquid hydrogen as fuel for space shuttles. The use of liquid hydrogen during a single space shuttle launch requires about 15,000 gallons per minute, which is equivalent to about forty-five hydrogen trailers, each with 13,000 gallons capacity. The hydrogen required to support a single Mars mission would be at least ten times that required for one space shuttle launch. In this work, we provide mass and energy balances, major equipment sizing, and costing of a hybrid CuO-CuSO4 plant with 1000 MW (≈ 30,240 kg/hr) H2 production capacity. With a 90% annual availability factor, the estimated hydrogen production rate is about 238,412 tons annually, the predicted plant efficiency is about 36%, and the estimated hydrogen production cost is about $4.0/kg (not including storage and transportation costs). In addition to hydrogen production, the proposed plant generates oxygen gas as a byproduct with an estimated flowrate of about 241,920 kg/hr (equivalent to ≈ 1,907,297 tons annually). We also propose a novel technology for separating SO2 and SO3 from O2 using a battery of redundant fixed-bed reactors containing CuO impregnated in porous alumina (Al2O3). This technology accommodates online regeneration of the CuO. Other practical approaches for gaseous separation are also examined including use of ceramic membranes, liquefaction, and regenerable wet scrubbing with slurried magnesium oxide or solutions of sodium salts such as sodium sulfite and sodium hydroxide. Finally, we discuss the applicability of high-temperature nuclear reactors as an ideal fit to providing thermal energy and electricity required for operating the hybrid thermochemical plant with high overall system efficiency.
A Development of Ceramics Cylinder Type Sulfuric Acid Decomposer for Thermo-Chemical Iodine-Sulfur Process Pilot Plant
Authors: Fukui Hiroshi , Minatsuki Isao , Ishino Kazuo
Published: September 2008

The hydrogen production method applying thermo-chemical Iodine-Sulfur process (IS process) which uses a nuclear high temperature gas cooled reactor is world widely greatly concerned from the view point of a combination as a clean method, free carbon dioxide in essence. In this process, it is essential a using ceramic material, especially SiC because a operation condition of this process is very corrosive due to a sulfuric acid atmosphere with high temperature and high pressure. In the IS process, a sulfuric acid decomposer is the key component which performs evaporating of sulfuric acid from liquid to gas and disassembling to SO2 gas. SiC was selected as ceramic material to apply for the sulfuric acid decomposer and a new type of binding material was also developed for SiC junction. This technology is expected to wide application not only for a sulfuric acid decomposer but also for various type components in this process. Process parameters were provided as design condition for the decomposer. The configuration of the sulfuric acid decomposer was studied, and a cylindrical tubes assembling type was selected. The advantage of this type is applicable for various type of components in the IS process due to manufacturing with using only simple shape part. A sulfuric acid decomposer was divided into two regions of the liquid and the gaseous phase of sulfuric acid. The thermal structural integrity analysis was studied for the liquid phase part. From the result of this analysis, it was investigated that the stress was below the strength of the breakdown probability 1/100,000 at any position, base material or junction part. The prototype model was manufactured, which was a ceramic portion in the liquid phase part, comparatively complicated configuration, of a sulfuric acid decomposer. The size of model was about 1.9m in height, 1.0m in width. Thirty-six cylinders including inlet and outlet nozzles were combined and each part article was joined using the new binder (slurry binder) and calcinated. Final polishing of the flange faces established in the entrance nozzles was also satisfactory. Many parts were joinable using new technology (new binder). For this reason, new technology is applicable to manufacture of not only a sulfuric acid decomposer but the instruments in the IS process, or other chemical processes.
Reformation of Regulatory Technical Standards for Nuclear Power Generation Equipments in Japan
Authors: Kurihara Mikio , Aoki Masahiro , Maruyama Yu , Takasaka Kiyosi , Nakajo Shigetada , Ogiso Zenichi , Goto Yukinori
Published: September 2008

Comprehensive reformation of the regulatory system has been introduced in Japan in order to apply recent technical progress in a timely manner. “The Technical Standards for Nuclear Power Generation Equipments”, known as the Ordinance No.622) of the Ministry of International Trade and Industry, which is used for detailed design, construction and operating stage of Nuclear Power Plants, was being modified to performance specifications with the consensus codes and standards being used as prescriptive specifications, in order to facilitate prompt review of the Ordinance with response to technological innovation. The activities on modification were performed by the Nuclear and Industrial Safety Agency (NISA), the regulatory body in Japan, with support of the Japan Nuclear Energy Safety Organization (JNES), a technical support organization. The revised Ordinance No.62 was issued on July 1, 2005 and is enforced from January 1 2006. During the period from the issuance to the enforcement, JNES carried out to prepare enforceable regulatory guide which complies with each provisions of the Ordinance No.62, and also made technical assessment to endorse the applicability of consensus codes and standards, in response to NISA’s request. Some consensus codes and standards were re-assessed since they were already used in regulatory review of the construction plan submitted by licensee. Other consensus codes and standards were newly assessed for endorsement. In case that proper consensus code or standards were not prepared, details of regulatory requirements were described in the regulatory guide as immediate measures. At the same time, appropriate standards developing bodies were requested to prepare those consensus code or standards. Supplementary note which provides background information on the modification, applicable examples etc. was prepared for convenience to the users of the Ordinance No. 62. This paper shows the activities on modification and the results, following the NISA’s presentation at ICONE-13 that introduced the framework of the performance specifications and the modification process of the Ordinance NO. 62.
Dismantling of Evaporators by Laser Cutting: Measurement of Secondary Emissions
Authors: Pilot Guy , Fauvel Sylvain , Gosse Xavier , de Dinechin Guillaume
Published: September 2008

In order to dismantle the evaporators of an obsolete reprocessing plant in Marcoule, studies were carried out by IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire)/DSU/SERAC in cooperation with CEA (power laser group) on the laser cutting of steel structures, on the request of COGEMA (now AREVA NC)/Marcoule (UP1 dismantling project manager) and CEA/UMODD (UP1 dismantling owner). The aim of these studies was: • to quantify and to characterize the secondary emissions produced by Nd-YAG laser cutting of Uranus 65 steel pieces representative of UP1 evaporator elements and to examine the influence of different parameters, • to qualify a prefiltration technique and particularly an electrostatic precipitator, • to compare the Nd-YAG used with other cutting tools previously studied. The experiments, which took place in a 35 m3 ventilated cutting cell, allow to underline the following points: • for the Uranus 65 steel, the sedimented dross, the deposits on the walls of the cutting cell and the aerosols drawn in the ventilation exhaust duct (∼ 275 m3/h), represent respectively between 92% and 99%, between 0.01% and 0.25% and between 1% and 8% of the total collected mass, • the attached slag varies much from one configuration to the other and can sometimes amount to a relatively important fraction of the total mass, • the kerves vary from 2 mm up to 7 mm for the Uranus 65 steel plates (thickness: 13.8 mm for the single plate and 12.8 + 3.5 mm for the double plate), • the exhausted aerosol mass per cut length (g/m) decreases with the cutting speed, varies neither with the stand-off nor with the gas pressure, is dependent upon the gas nature (for the double plate), increases with the laser power, is strongly affected by the nature of the steel (stainless steel or mild steel) and is independent upon the plate position, • the size distribution of aerosols is multimodal with a main mode often around 0.45 μm, • the electrostatic precipitator has been a satisfactory prefilter in the experiments with a filtration efficiency greater than 85% and the acoustic declogging system was very useful, • compared to other cutting tools already tested in similar conditions (reciprocating saw, 50 A plasma torch, 200 A plasma torch, grinder, 1 kW laser without assistant gas, arc-air and arc saw), the 4 kW laser produced less secondary emissions than all other tools and less aerosols than the other thermal tools.
Advanced In-Service Inspection Approaches Applied to the Phe´nix Fast Breeder Reactor
Authors: Guidez J. , Martin L. , Dupraz R.
Published: September 2008

The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960’s, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 °C. Inspection of the conical shell supporting the core diagrid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960’s, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant.
Modeling of Spacer Influence on Post-Dryout Heat Transfer in Heated Channels
Authors: Anglart Henryk
Published: September 2008

Post-dryout heat transfer plays an important role in safe and economical operations of Light Water Reactors (LWR). This type of heat transfer is avoided under normal operational conditions of nuclear reactors; however, it may occur in transient or accidential situations. To estimate the risk of clad damages due to increase of temperature associated with the occurrence of post-dryout, it is necessary to properly model heat transfer processes under such conditions. The influence of various parameters on heat transfer downstream of spacer has been investigated. It is concluded that heat transfer enhancement due to spacers is largely under-predicted for flows with relatively low quality. For such flows the effect of droplets impinging heated walls is significant and must properly be taken into account. The phenomenological model presented in this paper shows a superior accuracy over correlations and presents a potential to capture the phenomenon of rewetting that occurs downstream of spacers.
Fracture Behaviour of Type 304LN Stainless Steel and Its Welds
Authors: Dubey J. S. , Chakravartty J. K. , Singh P. K. , Banerjee S.
Published: September 2008

SA312 type 304LN stainless steel material, having closer control over impurities and inclusion content, is the intended piping material in the Advanced Heavy Water Reactors. Deformation, fatigue and fracture behaviour of this material and its weldments have been characterized at ambient temperature and at 558K. The details of the fractographic investigations and stretch zone width measurements are also discussed. The base metals shows high initiation toughness (>500 kJ/m2) and large tearing modulus at ambient and operating temperatures. Shielded Metal Arc Welding (SMAW) weld metal shows much much reduced initiation toughness and tearing resistance in comparison to base metal and Gas Tungsten Arc Welding (GTAW) welds. This is attributed to larger density of second phase inclusions in the SMAW weld metal. SZW measurements give a good alternate estimate of the toughness of the materials. Fatigue crack growth rate in SMAW weld metal was found to be comparable to base metal at higher load ratios.
Establishment of Korean Nuclear Ombudsman System Importing Compensation and Insurance Concept for Residents
Authors: Ko Hyun Seok , Kim Yong Min , Lee Young Wook , Shin Dong Hoon , Cho Young Ho , Kang Chang Sun
Published: September 2008

In Korea, the nuclear power generation is grown technically well. Already, 20 nuclear reactors are operated, and approximate they supply the 40% of whole the consumption of electric power. This is the driving force of Korean industrial development. Besides, Korean Standard Nuclear Power Plant that was developed by Korean self-technique with nuclear plant technique independence, Ul-Chin 6 has started the commercial operation. Advanced Korean Standard Nuclear Power Plant, new Gori 1, 2 constructions are commenced. But, past days Korean situation is that intention of residents is neglected in the decision making process of nuclear power plant construction and operation. In existing decision making process, it is regarded as the role of public opinion is secondary, and the problem of decision making process is that public is persuaded and believed. So, in decision making process, the public opinion is considered restrictively, there is not the actual public participation. Therefore the dissatisfaction of public is increased continuously, and in Korea, bad recognition about nuclear power is getting full. The method of public participation for complement of this problem is public hearing or ombudsman system. The public hearing is ensuring public participation before decision of a case, and ombudsman is the system that elevates the public satisfaction through continuous feedback of public requirement to occur in deciding and performing the matters. In Korean situation, that present 20 nuclear reactors are operated and also the place of radioactive waste repository has been decided, not only the introduction of public hearing to decide the coming matter but also the operation of ombudsman system to continuously correct and collect the public requirements about the matter to already decided and operated is necessary. In Korea, administration type ombudsman is operated now. But, it has operated without basic element at the aspect of organization, function and phase. So it is not established a firm phase as right relief body to be believed by public with lack of independence, authority and specialty. Therefore the establishment of organization that can be the role as special ombudsman organization about sensitive and special matter like nuclear matter is necessary. Definite establishment element of ombudsman system is to introduce of congress type ombudsman element, to be permanent standing system ombudsman, to limit jurisdiction extent and have strong authority, to be able to access easily, to be composed of legal and nuclear specialist. One of important requirements of resident is compensation. So, based on the resident compensation theory about aversion equipment, ombudsman system should be established introducing the insurance theory through risk management as functional background for appropriate compensation.
Rod Bundle Heat Transfer: Steady-State Steam Cooling Experiments
Authors: Spring J. P. , McLaughlin D. M.
Published: September 2008

Through the joint efforts of the Pennsylvania State University and the United States Nuclear Regulatory Commission, an experimental rod bundle heat transfer (RBHT) facility was designed and built. The rod bundle consists of a 7×7 square pitch array with spacer grids and geometry similar to that found in a modern pressurized water reactor. From this facility, a series of steady-state steam cooling experiments were performed. The bundle inlet Reynolds number was varied from 1 400 to 30 000 over a pressure range from 1.36 to 4 bars (20 to 60 psia). The bundle inlet steam temperature was controlled to be at saturation for the specified pressure and the fluid exit temperature exceeded 550 °C in the highest power tests. One important quantity of interest is the local convective heat transfer coefficient defined in terms of the local bulk mean temperature of the flow, local wall temperature, and heat flux. Steam temperatures were measured at the center of selected subchannels along the length of the bundle by traversing miniaturized thermocouples. Using an analogy between momentum and energy transport, a method was developed for relating the local subchannel centerline temperature measurement to the local bulk mean temperature. Wall temperatures were measured using internal thermocouples strategically placed along the length of each rod and the local wall heat flux was obtained from an inverse conduction program. The local heat transfer coefficient was calculated from the data at each rod thermocouple location. The local heat transfer coefficients calculated for locations where the flow was fully developed were compared against several published correlations. The Weisman and El-Genk correlations were found to agree best with the RBHT steam cooling data, especially over the range of turbulent Reynolds numbers. The effect of spacer grids on the heat transfer enhancement was also determined from instrumentation placed downstream of the spacer grid locations. The local heat transfer was found to be greatest at locations immediately downstream of the grid, and as the flow moved further downstream from the grid it became more developed, thus causing the heat transfer to diminish. The amount of heat transfer enhancement was found to depend not only on the spacer grid design, but also on the local Reynolds number. It was seen that decreasing Reynolds number leads to greater heat transfer enhancement.
Fuel Assembly Combined Lateral and Axial Model
Authors: Shah Sudhir J. , Brenneman Ben , Strumpell John H. , Williams Gary T.
Published: September 2008

The objective of this paper is to develop a purely mechanistic fuel assembly structural model that will predict the fuel assembly’s static and dynamic characteristics from the knowledge of the fuel assembly’s geometry and component properties. This model provides a method for analyzing the static and dynamic lateral and axial properties of the fuel assembly. A comparison of various in-air fuel assembly test data such as lateral and axial stiffnesses and lateral natural frequencies is provided to demonstrate the analytical model. The fuel assembly model developed by Shah, Brenneman, etc. (1), achieved very good agreement with assembly lateral impact test data by utilizing a “3-beam” model. In that model, the fuel rod-to-spacer grid interfaces were represented by spring and friction elements. The fuel assembly was restrained at each grid position by means of rotational springs, which were benchmarked to the test frequencies. This newly developed model eliminates the need for using rotational springs at the grid locations. Hence, it fully simulates the fuel assembly lateral and axial behavior based on the fuel assembly geometric properties. The fuel assembly model is a 2-D planar model of beams in both lateral and axial directions. The grids are modeled with plate elements. At each grid location there are springs, preload, and frictional sliders representing the lateral and axial connectivity characteristics to the fuel assembly beam model. As the Zircaloy grid preloads relax from irradiation, they can be easily simulated by removing the preload. Hence, this model can represent the fuel assembly structural properties for all aspects of fuel assembly cycles. This model can be used to analyze the fuel assembly lateral static stiffness, first mode and higher order lateral natural frequencies, mode shapes, axial stiffness, in-grid stiffness, through-grid stiffness, and fuel assembly lateral and axial seismic and LOCA response. The model will also estimate the fuel rod frequencies and mode shapes. This model may eliminate the need for some expensive prototype fuel assembly testing.
A New Developed Interface for CAD/MCNP Data Conversion
Authors: Shaaban Noha , Masuda Fukuzo , Nasif Hesham , Yamada Masao , Sawamura Hidenori , Morota Hidetsugu , Sato Satoshi , Iida Hiromasa , Nishitani Takeo
Published: September 2008

In a complex and huge system as in ITER fusion reactor, the creation of the geometrical input data of Monte Carlo (MC) codes such as MCNP is a highly exhausting task. Accordingly, it is a general approach to shift the geometric modeling into a computer aided design (CAD) system and to use an interface, which performs the exchange of CAD data into a representation appropriate for MC code. We have developed efficient algorithms and computer code, which are used to convert Parasolid format CAD files including solid and void data into MCNP input data. The CAD-MCNP conversion processes include creating surface equations; determining surface senses; constructing cell geometry and creating MCNP input file. This paper describes the basic algorithms used for the CAD/MCNP interface along with some applications for different geometries.
Validating CFD Models of Multiphase Mixing in the Waste Treatment Plant at the Hanford Site
Authors: Rosendall Brigette , Barringer Chris , Wen Feng , Knight Kelly J.
Published: September 2008

The Columbia River in Washington State is threatened by the radioactive legacy of the cold war. Two hundred thousand cubic meters (fifty-three million US gallons) of radioactive waste is stored in 177 underground tanks (60% of the Nation’s radioactive waste). A vast complex of waste treatment facilities is being built to convert this waste into stable glass (vitrification). The waste in these underground tanks is a combination of sludge, slurry, and liquid. The waste will be transported to a pre-treatment facility where it will be processed before vitrification. It is necessary to keep the solids in suspension during processing. The mixing devices selected for this task are known as pulse-jet mixers (PJMs). PJMs cyclically empty and refill with the contents of the vessel to keep it mixed. The transient operation of the PJMs has been proven successful in a number of applications, but needs additional evaluation to be proven effective for the slurries and requirements at the Waste Treatment Plant (WTP). Computational fluid dynamic (CFD) models of mixing vessels have been developed to demonstrate the ability of the PJMs to meet mixing criteria. Experimental studies have been performed to validate these models. These tests show good agreement with the transient multiphase CFD models developed for this engineering challenge.
Assessment of TRACE Code Using Rod Bundle Heat Transfer Mixture Level-Swell Tests
Authors: Welter Kent B. , Kelly Joseph M. , Bajorek Stephen M.
Published: September 2008

The TRAC/RELAP Advanced Computational Engine (TRACE) thermal-hydraulics code is currently under development by the United States Nuclear Regulatory Commission (NRC). TRACE is used for safety analyses of both conventional and advanced light water reactors. NRC assessed the prediction accuracy of the code by quantifying the axial void distribution in a rod bundle under low-pressure (0.16 to 0.44 Pa) and low-flow conditions (0.015 to 0.20 kg/s), using data obtained from the Rod Bundle Heat Transfer (RBHT) facility at Pennsylvania State University. NRC simulated 73 steady-state experiments (assessment cases) with variations in the total rod power, inlet subcooling, system pressure, and injection flow rate. Comparisons between TRACE calculations and RBHT data showed reasonable agreement. TRACE was found to over predict the bundle-exit void fraction by 13.3% with a linear goodness-of-fit (R2) of 0.87 and over predict the local void fraction by 10.1% with an R2 of 0.91. This paper discusses the models and correlations used in the TRACE calculation of mixture level swell, RBHT experimental results, modeling of the RBHT facility, and comparisons between data and code calculations.
Laser Isotope Enrichment for Medical and Industrial Applications
Authors: Eerkens Jeff W. , Kunze Jay F. , Bond Leonard J.
Published: September 2008

Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation repression. In this scheme, a gas of the selected isotopes for enrichment is irradiated with a laser at a particular wavelength that would excite only one of the isotopes. The entire gas is subjected to low temperatures sufficient to cause condensation on a cold surface or coagulation in the gas. Those molecules in the gas that the laser excited are not as likely to condense or dimerize (coagulate into a double molecule, called a dimer) as unexcited molecules. Hence in cold-wall condensation, gas drawn out of the system is enriched in the isotope that was laser-excited. We have evaluated the relative energy required in this process if applied on a commercial scale. We estimate the energy required for laser isotope enrichment is about 30% of that required in centrifuge separations, and 2% of that required by use of “calutrons”.
Study on the Accidental Rupture of Hot Leg or Surge Line in SBO Accident
Authors: Zhang Kun , Cao Xuewu
Published: September 2008

The postulated total station blackout accident (SBO) of PWR NPP with 600 MWe in China is analyzed as the base case using SCDAP/RELAP5 code. Then the hot leg or surge line are assumed to rupture before the lower head of Reactor Pressure Vessel (RPV) ruptures, and the progressions are analyzed in detail comparing with the base case. The results show that the accidental rupture of hot leg or surge line will greatly influence the progression of accident. The probability of hot leg or surge line rupture in intentional depressurization is also studied in this paper, which provides a suggestion to the development of Severe Accident Management Guidelines (SAMG).
The Application of Time-Related Event Tree During the HTR-PM Design Phase
Authors: Min Ping , Tong Jiejuan , He Xuhong
Published: September 2008

The Probabilistic Safety Assessment of HTR-PM, the demonstration nuclear power plant of High-Temperature Gas-Cooled Reactor in P.R. China, started in 2005, in order to offer supplementary support to the reactor design. Four critical initiating events families such as water-ingress, primary depressurization, loss of primary cooling etc. have been selected for the accident sequence analysis during the preliminary reactor design phase. Due to the special characteristics of doing PSA in the new reactor design phase, such as insufficient information, unavoidable iterations, complicated communications among multiple specialties and so on, efficient measures shall be developed to further the projects. A time-related Event Tree approach presented in this paper is one of them. Compared with ordinary event tree, the time-related event tree intends to illustrate not only the accident propagation processes, but also the key time points when safety systems, signals and operator actions are challenged and the durations. It seems to be a good bridge between designers and PSA engineers for the consistent understanding and more efficient information exchanging.