International Nuclear Information System (INIS)

The TVR heavy water research reactor was deployed at Moscow Institute of Theoretical and Experimental Physics. In 1990, the final batch of the spent nuclear fuel from this reactor was shipped to Production Association (PA) 'Mayak' for reprocessing. The SNF removal was a stage of the reactor decommissioning activities. The designs of the TVR reactor and its fuel elements are similar to the RA reactor designs. Two ways of the RA reactor SNF transportation to PA 'Mayak' have been considered: in aluminum barrels and in additional canisters using respectively TUK-32 and TUK-19 shipping casks. The practical experience and the equipment used to prepare for the TVR reactor SNF removal can be helpful to the RA reactor personnel in finding the best way to perform these engineering ...

Energy Technology Data Exchange (ETDEWEB)

JRR-3 is a research reactor of 10 MWt output, which attained the criticality in 1962. All the design, manufacture, installation and others of this reactor were carried out by Japanese technologies, except the fuel and heavy water as the moderator and coolant, therefore it is nicknamed Home-made No.1 Reactor. Recently, due to the change in the state of utilizing research reactors and the rise of quality in the utilization, JRR-3 has become to be unable to meet sufficiently the needs of users. The plan of reconstructing the JRR-3 was considered under such situation, and in order to reuse the reactor building, the reactor proper is removed, and an entirely new, high performance, versatile reactor is to be constructed. In this paper, as to the removal works of the JRR-3 ...

International Nuclear Information System (INIS)

... thermal power plants thermal reactors water cooled reactors WATER

International Nuclear Information System (INIS)

Eight years ago, corrosion and tube denting seriously threatened the reliability and design life of steam generators, especially for closed loop arrangements in pressurized water reactors (PWRs). Concentrated research by the Steam Generator Owners Group (SGOG) diagnosed the causes and produced effective solutions, notably guidelines for water chemistry control in the secondary loop. The guidelines recommend specific levels of water impurities and remedial actions to prevent cooling-water leaks in the condenser, prevent air leaks, limit corrosion product buildup, and remove some impurities while neutralizing others. Continued research in SGOB-II is investigating intergranular corrosion and stress corrosion cracking. 3 figures.

International Nuclear Information System (INIS)

Dr. Rudolf Schulten of West Germany and expert witnesses from national laboratories, utilities, and the nuclear industry testified on reactor safety issues as they relate to the Chernobyl accident and public concern that modern technology has not paid enough attention to public safety. Each of the witnesses contributed safety-related information based on what has been learned from the Soviet incident. Particular focus went to similarities and differences between the Chernobyl and US reactors in safety design and engineering and to the environmental effects of the accident. The N reactor near Richland and a commercial reactor at Fort St. Vrain, Colorado are the only two operating graphite reactors, but neither is a boiling water reactor.

Energy Technology Data Exchange (ETDEWEB)

The Materials and Components Technology Division (MCT) provides a research and development capability for the design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs related to nuclear energy support the development of the Integral Fast Reactor (IFR): life extension and accident analyses for light water reactors (LWRs); fuels development for research and test reactors; fusion reactor first-wall and blanket technology; and safe shipment of hazardous materials. MCT Conservation and Renewables programs include major efforts in high-temperature superconductivity, tribology, nondestructive evaluation (NDE), and thermal sciences. Fossil Energy Programs in MCT include materials development, NDE technology, and Instrumentation design. The division also has a complementary ...

International Nuclear Information System (INIS)

Most of all research reactors are immerged in the deep water pool to be a ultimate heat sink. At the neighbor of the reactor, some radio-active matters, such as Na-24, Ar-41, Mg-27, Al-28 and etc, may be generated by the neutron irradiation. Those radio-active isotopes may rise up to the pool water surface through the natural convection flow, which can make the radioactivity in the reactor hall rise high enough to concern about the health of people working in the reactor hall. When the irradiation test facilities are loaded or unloaded during a normal operation, the highly radio-activated primary coolant may flow out through the irradiation test holes on the top of the reactor. This also may be a main hazard source to make the working environment of the reactor hall bad. Making a hot ...

Energy Technology Data Exchange (ETDEWEB)

This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors.

International Nuclear Information System (INIS)

The development and successful application of a unique leak sealant formulation comprised of a mixture of graded, hollow ceramic microspheres, surface oxidized aluminum powder and saturated gibbsite suspension is described. The project was undertaken to address the escalating leakage from up to 15 small weld defects in the water reflector vessel, an integral component of the NRU (National Research Universal) reactor calandria. The reflector surrounds the reactor core with a neutron reflecting blanket of light water. Injection of the sealant is typically done with the reactor shutdown and the water reflector system operating normally, but can also be performed with the reactor at full power. The procedure is simple and effective. Individual treatments of as little as 125 ml of sealant (10 ppm in the 12,500 L system) have ...

International Nuclear Information System (INIS)

The direct injection of steam into a water pool is a method of heat transfer used in many process industries. The amount of research in this area however is limited to the nuclear industry, with applications relating to reactor cooling systems. Electrical resistance tomography (ERT), a low cost, non-invasive and which has high temporal resolution characteristics, can be used as a visualization tool for the resistivity distribution for the steam injection into water pool such as IRWST. In this paper, three dimensional resistivity distribution of the process is obtained through ERT using iterative Gauss-Newton method. Numerical experiments are performed by assuming different resistive objects in the water pool. Numerical results show that ERT is successful in estimating the resistivity distribution for the injection of steam in the water pool

Energy Technology Data Exchange (ETDEWEB)

The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides ...

Energy Technology Data Exchange (ETDEWEB)

An estimate of the tritium dose to the public in the vicinity of the heavy water research reactor facility at AECL-Chalk River Laboratories, Ontario, Canada, has largely been accomplished from analyses on regularly-collected samples of air, precipitation, drinking water and foodstuffs (pasture, fruit, vegetables and milk) and environmental dose models. To increase the confidence with which public doses are calculated, tritium doses were estimated directly from the ratio of tritiated species in urine samples from members of the general public. Single cumulative 24 h urine samples from a few adults living in the vicinity of the heavy-water research reactor facility at Chalk River Laboratories, Canada were collected and analysed for tritiated water and organically bound tritium. The participants were from Ottawa (200 km ...

Energy Technology Data Exchange (ETDEWEB)

The first Russian reprocessing facility, known as RT-1 (located at PA-Mayak in Ural region) was started on the radiochemical plant base in 1977. Nowadays RT-1 is the sole operating reprocessing plant in Russia. The main features of RT-1 is its broad spectrum of reprocessing spent nuclear fuels (SNF). The following spent fuel types are reprocessed: -) SNF from PWR reactors (WWER-440) and FB reactor (BN-600); -) SNF of transport ship reactors; -) production reactors SNF; -) research reactor spent nuclear fuel. The world-known technological processes are used at RT-1, but there are the following distinctive features. First, the universality of the three technological lines which allows not only the reprocessing of various SNF kinds, but also to implement the combined reprocessing of some types of them. Secondly, extraction of neptunium during SNF reprocessing which ...

International Nuclear Information System (INIS)

Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of ...

Energy Technology Data Exchange (ETDEWEB)

Information is presented concerning reactor research activities; isotope geology; NERC radiocarbon laboratory; teaching activities; and reactor operation.

International Nuclear Information System (INIS)

Recent achievements and tendency on reactor physics activities in Japan are reviewed according to topics published in journals or discussed at the Japan Research Committee on Reactor Physics.

International Nuclear Information System (INIS)

Optimal neutron energy for boron neutron capture therapy (BNCT) has been studied. Epithermal neutron is superior to thermal neutrons in treating deep-seated tumors. Design of the epithermal neutron column for BNCT has been performed by using a two-dimensional transport calculation code. Aluminum and heavy water are used as moderation materials. A thermal neutron column is also designed using heavy water as thermalization material. The configuration of the facility for treatment and research of BNCT and also for basic radio-biological studies of neutrons has been presented. (author).

International Nuclear Information System (INIS)

GRR-1 is a 5MW open pool type research reactor with MTR-type fuel elements cooled and moderated by light water with beryllium reflectors at the two opposing sides of the core. A graphite thermal neutron column is adjusted to one side of the core. Six radial horizontal beam tubes are available, of which three contain in-pile collimators for neutron scattering instruments. The reactor is currently out of operation for inspection and refurbishment purposes. The core has been dismantled and the fuel elements are stored in the used fuel storage tank. The GRR-1 inspection and refurbishment plan involves inspection and eventually replacement of the reactor's primary cooling circuit. The health physics procedures to be implemented during inspection of the main water outlet are divided in three stages: a) pool dose rate survey from pool top, b) pool drainage by ...

Energy Technology Data Exchange (ETDEWEB)

In nuclear reactor safety research, the question of the possible consequences of delayed core reflood during severe accidents or anticipated transient without scram transients in boiling water reactors (BWRs) has been raised. One can envisage a very low probability accident scenario leading to core uncovery and core heat-up, followed by control blade melting and subsequential delayed reflooding of the core with unborated water before its degradation. Reflooding of the hot core causes significant increases in the peak heating, melting, and hydrogen production rates, thus increasing the probability of core degradation. However, as has been established, debris beds formed from shattered fuel rods and quenched fuel melt will be undermoderated. The reflood process of a voided, intact core was examined using the TRAC/BFI CODE.

Energy Technology Data Exchange (ETDEWEB)

A two-phase flow regime map is developed experimentally and theoretically for a vertical hexagonal flow channel with and without a 36-finned rod hexagonal bundle. This type of flow channel is of interest to MAPLE-type nuclear research reactors. The flow regime maps are determined by visual observations and observation of waveforms shown by a capacitance-type void fraction meter. The experimental results show that the inclusion of the finned hexagonal bundle shifts the flow regime transition boundaries toward higher water flow rates. Existing flow regime maps based on pipe flow require slight modifications when applied to the hexagonal flow channel with and without a MAPLE-type finned hexagonal bundle. The proposed theoretical model agrees well with experimental results.

International Nuclear Information System (INIS)

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and ...

Energy Technology Data Exchange (ETDEWEB)

The Savannah River Swamp is a 3,020 Ha forested wetland on the floodplain of the Savannah River and is located on the Department of Energy`s Savannah River Site (SRS). Major impacts to the swamp hydrology occurred with the completion of the production reactors and one coal-fired powerhouse at the SRS in the early 1950`s. Water was pumped from the Savannah River, through secondary heat exchangers of the reactors, and discharged into three of the tributary streams that flow into the swamp. This continued from 1954 to 1988 at various levels. The sustained increases in water volume resulted in overflow of the original stream banks and the creation of additional floodplains. Accompanying this was considerable erosion of the original stream corridor and deposition of a deep silt layer on the newly formed delta. Heated water was discharged directly into Pen Branch and ...

Energy Technology Data Exchange (ETDEWEB)

The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor, based on the experienced light water reactor (LWR) technology, aiming at effective utilization of uranium resources, high burn-up and long operation cycle and plutonium multiple recycling. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional LWRs. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPC) in 1998, under technical cooperation with three Japanese reactor vendors. In the core design study of the RMWR, ...

International Nuclear Information System (INIS)

... 1343-3563 v. 57(1) SPECIFIC NUCLEAR REACTORS AND ASSOCIATED

Energy Technology Data Exchange (ETDEWEB)

The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant piping have led to excessive design costs, interference with normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National laboratory-sponsored research program aimed at investigating whether the probability of DEGB in Reactor Coolant Loop Piping of nuclear power plants is acceptably small such that the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimates the probability of indirect DEGB in Reactor Coolant piping as a consequence of seismic-induced structural failures within the containment of the GE supplied boiling water reactor at the ...

Energy Technology Data Exchange (ETDEWEB)

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of ...

Energy Technology Data Exchange (ETDEWEB)

The Fixed Bed Nuclear Reactor (FBNR) is a small 40 MWe reactor based on the Pressurized Water Reactor (PWR) technology. FBNR is an integrated primary circuit and simple in design. It has the characteristics of being small, modular, proliferation resistant, inherently safe and passively cooled reactor with reduced adverse environmental impact. It utilizes the fuel designed for high temperature reactors operating in a relatively low temperature of PWR environment The 15 mm diameter spherical fuel elements are made of TRISO type microspheres embedded in graphite and cladded by SiC. The coolant flow transfers them from the fuel chamber into the core and become fixed forming a suspended core. Any accident signal will cut off the power to the coolant pump causing a stop in the flow. This results in making the fuel elements fall out of the reactor ...

Energy Technology Data Exchange (ETDEWEB)

For light water reactors, loss of coolant is an important point in safety analysis, whereas for gas-cooled reactors the ingress of water into the core region is an incident of safety relevance. The applicability of the computer code system GAMTEREX to pebble beds of spherical high-temperature gas-cooled reactor fuel elements with simulated water ingress is verified by experiment. The measurements were performed at a Siemens-Argonaut reactor, using its ring core as a driver zone for a pebble-bed core in the center of the reactor.

Science.gov (United States)

In order to reduce the air concentration of (sup 3)H in the reactor buiIding of Wolsung Heavy Water Reactor, a computer code for estimation of adsorption behavior was programmed based on an equation derived for analysis of water vapor adsorption, and a ba...

Energy Technology Data Exchange (ETDEWEB)

In this paper the diffusion of tritiated water vapor into concrete walls is studied to evaluate tritiated water retention capacity of a fusion reactor concrete building. Using a model of the tritiated water diffusion determined form experimental results, depth profiles of tritiated water in concrete are calculated in the case of being exposed to air containing tritiated water vapor during the normal operational condition of a fusion reactor. A 0.5-m-thick concrete is sufficient for reactor hall walls from a viewpoint of the tritium containment.

Energy Technology Data Exchange (ETDEWEB)

The newly developed techniques by the Hot Laboratory (JMTR HL) have provided for us the key information on behavior of specimens due to mechanical / physical / chemical / synergistic effects of radiation, stress and water for fission and fusion reactor environment. These techniques are focused on several topics as follows; (1) miniaturized specimen test for the development of fusion reactor materials, (2) slow strain rate tensile testing (SSRT) and crack propagation measuring tests for the study of Irradiation Assisted Stress Corrosion Cracking (IASCC) of core internals of LWR, (3) handling technique on specimens including tritium for the research and development of tritium breeders and neutron multiplier as fusion blanket materials, (4) joining method using the Tungsten Inert Gas (TIG) welding technique for re-assembling of capsule and re-fabrication of specimen and (5) nondestructive evaluation using ...

Energy Technology Data Exchange (ETDEWEB)

Since 1976, the Nuclear Engineering Laboratory of the Technical Research Centre of Finland and Lappeenranta University of Technology have cooperated in the field of nuclear reactor thermal-hydraulics. During these years, a series of experimental facilities (REWET-I, -II, -III, VEERA) simulating pressurized water reactors (PWRs) have been built. The newest facility, PACTEL (Parallel Channel Test Loop), is an experimental out-of-pile facility designed to simulate the major components and system behaviour of a commercial PWR during postulated small and medium size break loss-of-coolant accidents (LOCAs), natural circulation and operational transients. A PACTEL natural circulation experiment has been carried out as an OECD/NEA international standard problem ISP 33. (2 refs., 3 figs., 2 tabs.).

Energy Technology Data Exchange (ETDEWEB)

The radiation control carried out in Atomic Energy Research Institute, Kinki University, for the reactor installation and the tracer/accelerator facilities from April, 1981, to March, 1982, is described. The reactor was operated for total 1057.1 hours at the maximum heat output of 1 W. The persons subject to radiation protection as of April, 1981, were 126 persons in all, including 23 in radiation work and 11 in X-ray work, etc. The contents of this report are as follows: personnel monitoring (health examination, the control of individual exposure dose); laboratory monitoring (the measurement of area dose rate, radioactive concentration in air and water, and surface contamination density); field monitoring (environmental ..gamma..-ray dose rate, radioactive concentration in environmental samples); the use of unsealed radioisotopes, etc.

International Nuclear Information System (INIS)

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used ...

International Nuclear Information System (INIS)

Results of ongoing research project at the McMaster Nuclear Reactor Facility on real-time neutron radiography for the visualization of interfacial geometry, movements and phase distributions in gas-liquid and gas-liquid-metal multi-phase flows are presented. Experiments were conducted with bubble column tubes with boiling liquid nitrogen, air-water and air-mercury mixtures. Discussions are also focused on air-water flowing within a tube containing a CANDU type 37 rod fuel bundle assembly positioned both horizontally and vertically. Computer processing using a digital image format to enhance the real-time images was used. Imaging techniques include frame averaging, background substraction, edge enhancement (spatial filtering), contrast enhancement and video densitometry. (orig.).

International Nuclear Information System (INIS)

Thermal hydraulics of a liquid metal is important to design the blanket of a magnetic confined fusion reactor. Since a liquid metal has high thermal and electrical conductivity, the flow characteristics are often different from those of an ordinary liquid like water especially in thermal convection and under a magnetic field. It is difficult to simulate such flows in a liquid metal cooled blanket by water. Flow visualization is a popular method to study thermal hydraulics. Since most of metals are visible by neutron rays, neutron radiography is available to the flow visualization of a liquid metal. The purpose of this study is to develop a visualization technique of the flow in a liquid metal by real-time neutron radiography using the tracer and the dye injection methods. A real-time thermal neutron radiography system of JRR-3M in Japan Atomic Energy Research Institute was used for the visualization ...

International Nuclear Information System (INIS)

Nuclear power in Romania was initiated on the basis of CANDU reactor type technology, an option found to fulfill the requirements for a sustainable economic development, to support the electric energy demand of the country and to ensure the population and environment protection. The construction of the Cernavoda NPP was heavily based on the Romanian industry participation and basic and applied nuclear research national resources. The experience acquired from Cernavoda NPP Unit 1 will be fructified in the construction of Units 2-5 to be built. The Romanian Ministry of Education and Research implemented a nuclear national program for research and development taking into account the European Union requirements and recommendations, the cooperation with the IAEA - Vienna and the Romanian government policy on short and medium terms in the nuclear field. The research-development program ...

Energy Technology Data Exchange (ETDEWEB)

THe high-temperature reactor is one of the alternatives for the now predominantly employed water-reactors. In a recently published book designer Rudolf Schulten outlines his concept. In this article the book is reviewed. (author). 1 ref.; 1 fig.

International Nuclear Information System (INIS)

Cooling water is sprayed on the outer surface of an upper portion of a container, and a pool is formed by the cooling water flowing down while cooling the container. Further, the cooling water stored in the cooling water pool is recycled by a pump for spraying the cooling water on the outer surface of the upper portion of the container. Sufficient amount of cooling water is supplied for spraying the cooling water to the outer surface of the upper portion of the container so that the outer surface of the container is free from drying and a liquid membrane is formed on the entire surface. The amount of the cooling water is made greater than that of the cooling water evaporated when the entire amount of the heat generate in the reactor core of the reactor is ...

Energy Technology Data Exchange (ETDEWEB)

Recommendations of the Advisory Committee on Reactor Safeguards are presented to the Commissioners for their consideration for FY 83 budget for the NRC safety research program.

International Nuclear Information System (INIS)

(1977). United States Parvez, A. Becker, M. Boguslawski, D. Harris,

Energy Technology Data Exchange (ETDEWEB)

We have been developing an innovative Steam Generator concept of Fast Breeder Reactors by using liquid-liquid direct contact heat transfer. In this concept, the SG shell is filled with a molten alloys, which is heated by primary sodium. Water is fed into the high temperature molten alloy, and evaporates by direct contact heating. In order to obtain the fundamental information to discuss the heat transfer mechanisms of the direct contact between the water and the alloy, this phenomenon was visualized by real-time neutron radiography. JRR-3M real-time thermal neutron radiography in Japan Atomic Energy Research Institute was used. Followings are main results. (1) The vigorous evaporation occurs in the molten alloy. This phenomena is different from the known phenomenon such as the evaporation of refrigerant R-113 in the water. (2) The evaporation in the bubble has finished in a moment ...

Energy Technology Data Exchange (ETDEWEB)

Korea Atomic Energy Research Institute (KAERI) has been developing the integral reactor. The reactor head structure assembly (RHSA) is the structure installed over the reactor cover. Due to the characteristics of an integral reactor, there are many instrument cables and power cables coming out from the reactor cover and main components. The RHSA provides an interface location to connect these cables from Architecture Engineer (AE) and System Designer (SD). It also prevents a pipe whip and it prohibits instruments from becoming missiles. In this research, the axiomatic design approach for the RHSA is performed.

International Nuclear Information System (INIS)

(Jun 1982). United States Davis, MV Berger, H. Patricelli, F. Georgia Institute

International Nuclear Information System (INIS)

Wolsung nuclear power plant has experienced four occasions of reactor shutdown owing to heavy water leaks since its commercial operation. Among these heavy water leaks, only one case was acute and brought about reactor shutdown but the other cases listed below were chronic and repaired after manual reactor shutdown. (author). 4 tabs., 10 figs.

International Nuclear Information System (INIS)

The Fact Finding Committee on 'Future Nuclear Power Policy' established by the 8th German Federal Parliament in its report of June 1980 among other items published the recommendation to commission a 'risk oriented analysis' of the SNR-300 in order to enable a pragmatic comparison to be made of the safety of the German prototype fast breeder reactor and a modern light water reactor (a Biblis B PWR). The Federal Minister for Research and Technology in August 1981 officially commissioned the Gesellschaft fuer Reaktorsicherheit (GRS) to conduct the study. Following a recommendation by the Fact Finding Committee, additional studies were performed also by a group of opponents of the breeder reactor. On the instigation of the group of opponents the delivery date of the study was altered several times and finally set at April 30, 1982. GRS submitted its report by this deadline. However, a ...

Energy Technology Data Exchange (ETDEWEB)

Upon request by the Science and Technology Agency of Japanese Government, the Japan Atomic Energy Research Institute has conducted Piping Reliability Test Program to demonstrate the safety and reliability of light water reactor primary pipings. In this report, the results of the program are summarized. In the test program, pipe fatigue tests, Leak-Before-Break (LBB) verification tests and pipe rupture tests were carried out to examine the integrity of pipings, to verify the LBB concept and to demonstrate the effectiveness of the protective measures against jet impingement and pipe whip under pipe rupture event, respectively. In the pipe fatigue tests, a procedure to predict the fatigue crack growth was developed and the integrity of piping during plant service life was demonstrated. In the LBB verification tests, pipe fracture tests and leak rate tests were performed using cracked pipes. Based on the test results, LBB in ...

International Nuclear Information System (INIS)

The steam generator low water level trip setpoint of KORI NPP units 3 and 4(KNU 3 and 4), three-loop Westinghouse pressurized water reactor, is higher than that of OPR1000. In addition, steam generator downcomer water level in KNU 3 and 4 could fluctuate easily during a transient because of smaller downcomer water inventory, compared to the total water inventory in the steam generator. Due to these reasons, there is a higher possibility of unnecessary reactor trips caused by the steam generator low-low water level in KNU 3 and 4. Its operating history shows that most of reactor trips were caused by steam generator low-low level reactor trip signal. Such reactor trips, especially unnecessary ones, result in time and economic losses. In this paper, a ...

International Nuclear Information System (INIS)

The results for development of methods and computer programs for integrated optimization of parameters of perspective fast reactors are given. The possibilities of the program for the reactor campaign calculation are analysed. This program is based on utilisation of the Bubnov-Galerkin method and Wigner disturbance theory. The possibility of application of approximation methods for the optimization researches is discussed. The results of development of the programs for complex reactor computations with account of control rods system and change of physical parameters in the reactor campaign are discussed. (author).

Energy Technology Data Exchange (ETDEWEB)

The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully ...

Science.gov (United States)

Utility interest has recently increased in potential future nuclear units that combine the characteristics of smaller size, greater simplicity, and more passive safety features. In response to such interest, General Electric (GE) began development in 1982 of a 600-MW(electric) reactor with simplified power generation and safety systems. This paper provides an overview of the simplified boiling water reactor (SBWR) design, with emphasis on the thermal-hydraulic aspects of the design. The SBWR is a natural circulation reactor requiring no pumps to circulate the water through the core.

Science.gov (United States)

average bulk water temper ature rise, OF bulk water temperature at elevation z, OF bulk water temperature in channels 0 and 1, O F film temperature, OF ...

International Nuclear Information System (INIS)

Decommissioning of radiological and nuclear installations is for this century the new challenge. One of the performance criteria is the reduction of total quantities of radioactive materials (liquid or solid) arising from dismantling and decontamination of radiological and nuclear installations. In this work we present a new application of the water soluble polymers used as: - flocculation agents in treatment and conditioning process within the management of radioactive liquid materials; - strippable coatings on solid materials based on the water soluble polymers. The parameters of water soluble polymers made in our Institute by radiation processing have been analysed, namely the molecular average weight, composition, and efficiency of utilization of these polymeric materials as well as the content of ash, additives, decontamination factor, consumption per surfaces/liter, corrosion aspects, compatibility with various ...

Energy Technology Data Exchange (ETDEWEB)

A method for determining the reactivity of highly subcritical systems of fissile material, using neutron-noise power spectral densities in conjunction with a /sup 252/Cf source, had previousy been tested in two fast reactor critical assemblies (a mockup of the Fast Flux Test Facility reactor and unreflected enriched uranium metal assemblies) and one thermal reactor (a light-water moderated and reflected lattice of Oak Ridge Research Reactor fuel elements). The last-mentioned test demonstrated the effectiveness of the method in water-moderated systems and thereby prompted the present study of its application to facilities for fuel preparation, reprocessing, and storage. To investigate the applicability of this method to facilities for fuel preparation, reprocessing, and storage, limited experiments were performed with a uranyl fluoride ...

Energy Technology Data Exchange (ETDEWEB)

Among the several chemical processes conducted for the removal of organic matter present in wastewaters coming from some agro-industrial plants (wine distilleries, olive oil mills, etc), the oxidation by ozone has shown a great effectiveness in the destruction of specially refractory pollutants: it is demonstrated that the biodegradability of those wastewaters increases aflcer an ozonation pretreatment. Their great pollutant character is imputed to the presence of some organic compounds, like phenols and polyphenols, which are toxic and inhibit the latter biological treatments. In this research, a competitive kinetic procedure reported by Clurol and Nekouinaini is applied to determine the degradation rate constants by ozone of several phenolic acids which are present in the wastewaters from the olive oil obtaining process. The resulting kinetic expressions for the ozonation reactions are useful for the successful design and operation of ozone ...

Energy Technology Data Exchange (ETDEWEB)

The Fast Flux Test Facility (FFTF) is a 400 Mw(t), loop-type, sodium-cooled, fast neutron reactor. It is operated by the Westinghouse Hanford Company for the United States Department of Energy at Richland, Washington. The FFTF is a multipurpose test reactor used to irradiate fuels and materials for programs such as Liquid Metal Reactor (LMR) research, fusion research, space power systems, isotope production and international research. FFTF is also used for testing concepts to be used in Advanced Reactors which will be designed to maximize passive safety features and not require complex shutdown systems to assure safe shutdown and heat removal. The FFTF also provides experience in the operation and maintenance of a reactor having prototypic components and systems typical of large LMR (LMFBR) power plants. The 5 year ...

International Nuclear Information System (INIS)

Indian Nuclear Power Programme is based on closed nuclear fuel cycle for efficient utilization of its nuclear resources. This strategy also enables waste classification and gives an elegant solution to long-lived waste disposal problem. The three stage nuclear programme envisages mainly pressurized heavy water reactors in the first stage, fast breeder reactors in the second stage and thorium utilization in the third stage. Advanced Fuels in the context of this paper refer to Pu bearing fuels used or proposed to be used in our three stage programme. Fabrication of (U-Pu) Mixed Carbide fuel for FBTR is carried out at Radio Metallurgy Division at Trombay which has also an excellent Characterization facility required for development of all types of advanced Fuels. A (U-Pu) MOX fuel required for Proto-type Fast Breeder Reactor (PFBR-500 MWe) is carried out at Advanced Fuel Fabrication Facility (AFFF), ...

Energy Technology Data Exchange (ETDEWEB)

The days of high-temperature reactors in the Federal Republic of Germany are numbered. The AVR has been decommissioned, and an application has been filed for licensing the decommissioning of the THTR. Nevertheless, Prof. Dr. Rudolf Schulten who is the director of Juelich Nuclear Research Center's Institute for Reactor Development, and also full professor of Aachen Technical University in the field of reactor safety, predicts a good future for the HTR reactor line on a worldwide level, due to the inherent safety of this reactor type. (orig.).

International Nuclear Information System (INIS)

Full text of publication follows: The heat transfer and flow in narrow channels has lots of advantages such as compact structure, high efficiency, design flexibility and so on. So it is widely used in the fields such as the new reactor core plate elements, the once-through stream generator, compact heat exchangers as well as electronic components. In recent years, more strong attentions have been attracted to the thermal-hydraulic characteristics and mechanism of the two-phase flow in narrow channels. As the flow regime characteristics of two-phase flow is fundamental one of them, the research on the two-phase flow regimes and the regime transitions in horizontal rectangular narrow heated channels can provide theoretical foundation and engineering directions to the whole research on the thermal-hydraulic characteristics and mechanism of the two-phase flow in narrow channels. The characteristics of two-phase flow regimes and ...

Energy Technology Data Exchange (ETDEWEB)

The most relevant aspects of research and service experience with environmentally-assisted cracking (EAC) of carbon (C) and low-alloy steels (LAS) in high-temperature (HT) water are reviewed, with special emphasis on the primary pressure boundary components of boiling water reactors (BWRs). The main factors controlling the susceptibility to EAC under light water reactor (LWR) conditions are discussed with respect to crack initiation and crack growth. The adequacy and conservatism of the current BWRVIP-60 stress corrosion cracking (SCC) disposition lines (DLs), ASME III fatigue design curves, and ASME XI reference fatigue crack growth curves, as well as of the GE EAC crack growth model are evaluated in the context of recent research results. The operating experience is summarized and compared to the experimental/mechanistic background ...

Energy Technology Data Exchange (ETDEWEB)

We defined the Korean unique safety-critical software development methodology by modifying Dr. Harel`s statechart-based on formal methods in order to digitalized the reactor protection system. It is suggested software requirement specification guideline to specify design specification which is basis for requirement specification and automatic programming by the caused by shutdown parameter logic of the steam generator water level for Wolsung 2/3/4 unit SDS no.1 and simulated it by binding the Graphic User Interface (GUI). We generated the K and R C code automatically by utilizing the Statemate MAGNUM Sharpshooter/C code generator. Auto-generated K and R C code is machine independent code and has high productivity, quality and provability. The following are the summaries of major research and development. - Set up the Korean unique safety-critical software development methodology - Developed software requirement ...

International Nuclear Information System (INIS)

To obtain reactor physics parameters for high-moderation mixed-oxide (MOX) cores, Nuclear Power Engineering Corporation (NPEC), the French Atomic Commission (CEA), and their industrial partners have conducted a MOX core physics experimental program called MISTRAL with the EOLE critical facility of the Cadarache research center. This program consists of four high-moderation cores and was successfully completed in July 2000. This paper describes the analysis results of MISTRAL-3 that is a homogeneous full MOX cylindrical core (H/HM = 6.2) with an 80-cm height and a 59-cm diameter consisting of 1388 standard pressurized water reactor-type MOX fuel rods of 7.0 wt% plutonium-enrichment in a square pitch of 1.39 cm. NPEC has been analyzing the experimental results by using the SRAC and MVP code systems. SRAC and MVP calculate the nuclear core characteristics correctly for the high-moderation MOX core ...

International Nuclear Information System (INIS)

Presently, industrial maturity can be claimed for two fuel cycle strategies, viz. the 'Once Through Fuel Cycle' (OTC), and the 'Reprocessing Fuel Cycle' (RFC) in which plutonium and very limited uranium quantities are being recycled. It is helpful to recall some key data that set the stage for any discussion of fuel cycle options: 1. Worldwide, the annual spent fuel discharge is in the range of 10500-11000 t heavy-metal (HM), while the industrial reprocessing capacity amounts to #approx# 5000 t HM (OECD NUCLEAR ENERGY AGENCY, Accelerator-driven Systems (ADS) and Fast Reactors (FR) in Advanced Nuclear Fuel Cycles: a Comparative Study, Paris, 2002). Hence, less than 1/2 of the discharged spent fuel can be processed. 2. Worldwide, the cumulative inventory of stored spent fuel is estimated to be #approx# 190000 t HM, and the amount of reprocessed spent fuel is estimated to be #approx# 70000 t HM. The latter inventory has been transformed into high-level waste (HLW) and ...

Science.gov (United States)

The calculated radioactive waste inventories of the Turkey Point pressurized water fission reactor (PWR) and the Starfire conceptual fusion tokamak are compared as a function of time from initial start-up to 10,000 years after decommissioning. Only materi...

Science.gov (United States)

This is the final Report for the technical ''on-the-job traning'' for the Wolsung CANDU nuclear power plant which is the first Pressurized Heavy Water Reactor setting up in Korea. The technical ''on-the-job traning'' was established to increase the capabi...

Energy Technology Data Exchange (ETDEWEB)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens ...

Energy Technology Data Exchange (ETDEWEB)

The objective of this research was to determine the extent of damage that occurs when two pipes experience an impact event due to one whipping against the other. The research was conducted through experimental and analytical approaches. The former required the development of a specialized impact machine that could accelerate a whipping pipe with sufficient energy to cause failure of a target pipe that was heated and pressurized to Pressurized Water Reactor (PWR) conditions. Damage was measured in terms of crushing, bending, and failure. The results of the tests permitted the correlation between pipes of a certain size and the damage they could cause when impacting with a certain amount of known energy. These results were used to evaluate the pipe whip criteria in the Standard Review Plan 3.6.2-4. It was established that the criteria conditions did not fully represent the results obtained experimentally. ...

Science.gov (United States)

developing a consensus PRA standard for non- associated guidance light water reactor applications that will address some aspects of PRA in design. ...

International Nuclear Information System (INIS)

Reduced-Moderation Water reactor (RMWR) is a light water breeder reactor developed by Japan Atomic Energy Research Institute (JAERI). The RMWR comprises tight lattice fuel assemblies with gap clearance of around 1.0 mm to reduce water volume ratio to achieve a high conversion ratio. It is important to estimate the thermal hydraulic safety margin of the tight lattice core of the RMWR. In the present study, the boiling transition (BT) prediction performance of the subchannel analysis code NASCA developed for the current BWR cores was assessed for series of tight lattice critical heat flux (CHF) experiments performed in JAERI. The test section was a 7-rod bundle with rod diameter of 12.3 mm, rod gap of 1.0 mm and heated length of 1.8m. Axial power distribution was flat. With a simple subchannel model, the code overestimates the critical power in the high mass ...

International Nuclear Information System (INIS)

Since the 1970's, global efforts have been going on to replace the high-enriched (>90% "2"3"5U), low-density UAlx research reactor fuel with high-density, low enriched (<20% "2"3"5U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U_3Si_2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher ...

International Nuclear Information System (INIS)

In a supercritical water-cooled reactor, property of water changes significantly around the critical point. It is expected that irradiation and change of water property will affect the chemistry and material corrosion. Deep understanding of interactions between supercritical water and materials under irradiation is important. However, comprehensive data on radiolysis, kinetics, corrosion and thermodynamics have not been obtained due to the severe experimental condition. To get such data by experiments and computer simulations, a national program funded by Ministry of Education, Culture, Sports, Science and Technology (MEXT) has been started since December 2002. (author)

Energy Technology Data Exchange (ETDEWEB)

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

International Nuclear Information System (INIS)

Manganese is a common contaminant of mine water and other waste waters. Due to its high solubility over a wide pH range, it is notoriously difficult to remove from contaminated waters. Previous systems that effectively remove Mn from mine waters have involved oxidising the soluble Mn(II) species at an elevated pH using substrates such as limestone and dolomites. However it is currently unclear what effect the substrate type has upon abiotic Mn removal compared to biotic removal by in situ micro-organisms (biofilms). In order to investigate the relationship between substrate type, Mn precipitation and the biofilm community, net-alkaline Mn-contaminated mine water was treated in reactors containing one of the pure materials: dolomite, limestone, magnesite and quartzite. Mine water chemistry and Mn removal rates were monitored over a 3-month ...

Energy Technology Data Exchange (ETDEWEB)

This report discusses a desk-top computer program has been developed for estimating the costs, waste volumes, and occupational radiation exposures associated with decommissioning light-water reactor power stations. Cost categories and cost algorithms used in the program are discussed and a brief description of the user interface is given.

International Nuclear Information System (INIS)

Purpose: To eliminate the water level deviation due to the recycling flowrate, as well as enable a stable control to a reference value even upon changes in the recycling flowrate caused by the variation in the opening degree of a minimum flow valve. Constitution: Reactor recycling system comprises a feedwater pump, a flowrate control valve, a reactor water level detector, and a minimum flow line and a minimum flow valve for preventing the overheating of the feedwater pump at a low flowrate. A flowrate compensator is further disposed, in which a recycling flowrate signal is subtracted from a pump flow rate signal and the result is fedback as a compensated pump flowrate signal. This enables the control system to operate at a rapid response for suppressing the effect of the recycling flowrate as external disturbance, whereby the water level in the reactor can be ...

International Nuclear Information System (INIS)

The European Research and Development Program on decay heat removal by natural convection for the European Fast Reactor (EFR) covers the calculational methods and the model experiments performed for code validation. The studies concentrate on important physical effects of the cooling modes within the primary system and the direct reactor cooling circuits and include fundamental tests as well as reactor experiments. (author)

Energy Technology Data Exchange (ETDEWEB)

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water - 310degC and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been ...

Science.gov (United States)

The long operating cycle simplified boiling water reactor is a reactor concept that pursues both safety and the economy by employing a natural circulation reactor core without a refueling, a passive decay heat removal, and an integrated building for the reactor and turbine. Throughout the entire spectrum of the design basis accident, the reactor core is kept covered by the passive emergency core cooling system. The decay heat is removed by the conventional active low-pressure residual heat removal system. As for a postulated severe accident, the suppression pool water floods the lower part of the reactor pressure vessel (RPV) in the case when core damage occurs, and the in-vessel retention that keeps the melt inside the RPV is achieved by supplying the coolant. The containment adopts a parallel-double-steel-plate ...

International Nuclear Information System (INIS)

Active Process Water (APW) System, provided as unitized closed loop system in Tarapur Atomic Power Project Units-3 and 4, serves to remove heat from various heat exchangers. One of the important loads served by APW system is shutdown cooling heat exchangers and if APW shutdown is taken then reactor cannot be maintained in cold shutdown condition. It is estimated that after 7 days of reactor shutdown, if about 20% of the normal cooling flow to shutdown cooling heat exchangers is provided then along with keeping PHT in cold shutdown state, reactor components, moderator, end shield water, calandria vault water and calandria vault concrete temperature can be maintained within technical specification limits for extended duration. (author)

Energy Technology Data Exchange (ETDEWEB)

Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, ...

Energy Technology Data Exchange (ETDEWEB)

This patent describes a nuclear system comprising a containment airspace in which a nuclear reactor pressure vessel is disposed there being a reactor core within the pressure vessel. It comprises a heat exchanger elevated a distance above the pressure vessel; a pool of water surrounding the heat exchanger; means for venting the pool of water to an environment outside the containment; a heat exchanger entry conduit within the containment, the entry conduit having an open lower end communicating with the containment space, and an upper end connected to the heat exchanger, water-containing heated fluid present in the containment airspace incident a pressure vessel loss of coolant event entering and flowing through the entry conduit into the heat exchanger for cooling the fluid to convert water vapor therein to a condensate and separate non-condensable gasses ...

International Nuclear Information System (INIS)

Full text: Full text: The incomplete understanding of the complex mechanisms connected with the interaction between thermal-hydraulic and neutron kinetics still challenges the design and the operation of nuclear reactors and imposes the adoption of conservatism in the evaluation of safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience suggests the revisiting of those areas and the identification of design/operation requirements that can be relaxed. So far, almost all of the safety analyses of research reactors have been performed using conservative computational tools such as channel codes but, nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity. The global aim of the current work is an attempt to apply the best-estimate system thermal-hydraulic code Relap5. For this purpose, the generic IAEA ...

International Nuclear Information System (INIS)

Since January, 1981, the project of development of nuclear fuel fabrication technology for Wolsung reactor (CANDU type) was undertaken by KAERI(Korea Advanced Energy Research Institute) and successfully fulfilled with loading 24 fuel bundles made by KAERI in Wolsung reactor in September, 1984. On the basis of this accumulated technology and experience, mass production plan to supply all the nuclear fuels for Wolsung reactor is under way. In this presentation, the Korean experience in the development of the nuclear fuel fabrication technology, safety and performance evaluation of KAERI fuel and the results of irradiation of KAERI fuels in Wolsung reactor will be described.

International Nuclear Information System (INIS)

SCDAP/RELAP5 code has been developed in US for best-estimate simulation of light water reactors transients during nuclear accidents. The code models the coupled behaviour of the cooling system, reactor core and fission products release during the accident. It is the result of the coupling between RELAP5, modelling thermal hydraulic, control system, reactor kinetics and the transport of noncondensable gases, and SCDAP code modelling the behaviour of the reactor core during severe accidents. The paper briefly presents the application of SCDAP/RELAP5 code to CANDU severe accident analysis. Also, the paper proposes a summary of the needs for development that could enhance the quality of the severe accidents related predictions in CANDU reactors. (authors)

Energy Technology Data Exchange (ETDEWEB)

The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. ...

Wastenet

...INFORMATION Diarrhoeal disease Food safety and foodborne illness Enterohaemorrhagic Escherichia coli (EHEC) Cholera WHO PROGRAMMES AND ACTIVITIES Child and Adolescent Health and Development (...FOS) Initiative for Vaccine Research (IVR) TECHNICAL INFORMATION Vaccine research: diarrhoeal diseases Cholera Water-related diseases Household water treatment and safe storage WHO Global Salm-Surv ...PUBLICATIONS Diarrhoea: child and adolescent health Diarrhoea: cholera RELATED TOPICS - Child health - Water - Food safety - Cholera - Travel - Breastfeeding ...

Energy Technology Data Exchange (ETDEWEB)

The purpose of this study is to investigate the feasibility of visualization and void fraction measurement of air-water two-phase flow in a small diameter tube (I.D.: 4.08 mm) by using the real-time neutron radiography and image processing techniques. Video images of two-phase flow were taken by using the real-time neutron radiography system (thermal neutron radiography facility No.2) installed at the Japan Research Reactor 3M of the Japan Atomic Energy Research Institute. The shape of bubbles and its moving behavior were clearly observed from the video images. The image corrections for dark current, shading, field intensity fluctuation and electrical system drift were examined in order to measure the void fraction from the video images. Though, generally speaking, the effect of the scattered neutron could not be ignored for quantification of the images taken by the neutron radiography, the scattered ...

International Nuclear Information System (INIS)

The purpose of this study is to investigate the feasibility of visualization and void fraction measurement of air-water two-phase flow in a small diameter tube (I.D.: 4.08 mm) by using the real-time neutron radiography and image processing techniques. Video images of two-phase flow were taken by using the real-time neutron radiography system (thermal neutron radiography facility No.2) installed at the Japan Research Reactor 3M of the Japan Atomic Energy Research Institute. The shape of bubbles and its moving behavior were clearly observed from the video images. The image corrections for dark current, shading, field intensity fluctuation and electrical system drift were examined in order to measure the void fraction from the video images. Though, generally speaking, the effect of the scattered neutron could not be ignored for quantification of the images taken by the neutron radiography, the scattered ...

International Nuclear Information System (INIS)

Current light water reactor (LWR) steam generators have been affected by a variety of corrosion and mechanical damage degradation mechanisms. Included are wear caused by tube vibration, intergranular corrosion, pitting, and thinning or wastage of the steam generator tubing and accelerated corrosion of carbon steel supports (denting). The Electric Power Research Institute (EPRI) and the Steam Generator Owners Groups (I, II) have sponsored laboratory and field studies to provide ameliorative actions for the majority of the damage forms experienced to date. Some of the current corrosion mechanisms are aggravated or caused by unique materials choices or materials interactions. New materials have been proposed and at least partially qualified for use in replacement model steam generators, including an advanced LWR design. In so far as possible, the materials choices for the advanced LWR steam generator avoid the corrosion ...

International Nuclear Information System (INIS)

The release rate of a nuclide from a reactor or a radioactive waste disposal plant at the accident is not steady, but varies with time. The various parameters of a nuclide migration into environment vary also day after day, or with the seasons. In such cases, dynamic behavior of the nuclide in the environment must be taken into consideration. It is difficult for a mathematical model to involve all of mechanisms for the nuclide migration. The environment for evaluation of doses are usually divided into some of compartments in which a nuclide concentration is uniform. Time variations of the nuclide concentration in the compartment are described in simultaneous differential equations. The nuclide concentration can be solved as a time function, and the radiation doses, therefore, can be estimated as a time function. Generic analysis code for dynamic compartment model (GACOM) is developed for the nuclide migration and the evaluation of doses in terrestrial biosphere. ...

International Nuclear Information System (INIS)

The application of chemical cleaning for dissolving and removing scale and sludge is being planned in the Japanese pressurized water reactor (PWR) plant in order to maintain high heat transfer performance and to prevent steam generator (SG) tube degradation. In this paper, the effectiveness of the Electric Power Research Institute (EPRI) and German Kraftwerk Union (KWU) processes on the integrity of structural materials other than SG tubes and the comprehensive applicability of chemical cleaning are discussed. The integrity of structural materials such as carbon steel, low-alloy steel and stainless steel was maintained after the EPRI and KWU processes. KWU chemical cleaning tailored for crevice cleaning has been studied to improve its cleaning effectiveness in crevices and to control the corrosion depth of structural materials less than the criterion for corrosion depth. (author)

Energy Technology Data Exchange (ETDEWEB)

This report summarizes the major activities and accomplishments of the Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Radiotherapy Research Program for calendar year 2001. Applications of supportive research and development, as well as technology deployment in the fields of chemistry, radiation physics and dosimetry, and neutron source design and demonstration are described. Contributions in the fields of physics and biophysics include development of advanced patient treatment planning software, feasibility studies of accelerator neutron source technology for Neutron Capture Therapy (NCT), and completion of major modifications to the research reactor at Washington State University to produce an epithermal-neutron beam for NCT research applications.

International Nuclear Information System (INIS)

In an existent emergency reactor core cooling device, if a ruptures should occure in a pipeline of a gravitational dropping type reactor core cooling system pool (GDCS) due to some or other causes, a portion of GDCS pool water was flown out of the ruptured port and could not be used for reactor core cooling. Then, a difference pressure detector is disposed to a GDCS pipeline at the inlet of a reactor pressure vessel. When it is judged by the detector, that coolants flow to the outside of the injection pipeline, an injection value disposed to the GDCS pipeline is closed by the difference pressure signal. Even if a rupture should occur on the side of the pressure vessel at downstream to the check value of the GDCS pipeline, since backflow is caused at the pressure container inlet of the GDCS pipeline with the rupture port, the rupture is detected by the difference pressure detector to ...

British Library Electronic Table of Contents (United Kingdom)

An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25mx0.05m) and 2.59m, respectively, whereas the inclination angle of the riser is 50degree. The flow was captured by a high-speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Countercurrent flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels mea...

Science.gov (United States)

In the past few decades the need for improved nuclear reactor safety analyses has led to a rapid development of advanced methods for multidimensional thermal-hydraulic analyses. These methods have become progressively more complex in order to account for the many physical phenomena anticipated during steady state and transient Light Water Reactor (LWR) conditions. The advanced thermal-hydraulic subchannel code COBRA-TF (Thurgood, M. J. et al., 1983) is used worldwide for best-estimate evaluations of the nuclear reactor safety margins. In the framework of a joint research project between the Pennsylvania State University (PSU) and AREVA NP GmbH, the theoretical models and numerics of COBRA-TF have been improved. Under the name F-COBRA-TF, the code has been subjected to an extensive verification and validation program and has been applied to variety of LWR steady state and transient ...

Energy Technology Data Exchange (ETDEWEB)

This is the seventh report by the Advisory Committee on Reactor Safeguards (ACRS) in response to the Congressional requirement for an annual report on the Nuclear Regulatory Commission (NRC) Reactor Safety Research Program. As previously requested by the Congress, the timing of this report has been adjusted to enable the ACRS to address the proposed budget for FY 1985 that has been submitted to the Congress by the President. Part I is a compilation of our comments and recommendations regarding the NRC Safety Research Program budget for FY 1985. It is intended to serve as an Executive Summary. Part II is divided into eight chapters, each of which represents a Decision Unit of the NRC research program. In each chapter, specific comments are included on the research involved in the Decision Unit, an assessment of priorities, and recommendations regarding new ...

International Nuclear Information System (INIS)

The neutron radiography facility was installed at the tangential beam port of the 3 MW TRIGA MARK-II research reactor. In the facility only direct film neutron radiography method is being used. The project involves development of electronic imaging system for real time neutron radiography in the existing facility with the aim of utilizing it for research and industrial applications. In establishing the electronic imaging system for real time neutron radiography the improvements of existing facility were almost done during this period. In parallel, the former facility was used for the research: (a) A study of wood and wood plastic composites with and without additive by using film neutron radiography and (b) A study of jute reinforced polymer composites by using film neutron radiography technique. (author)

Energy Technology Data Exchange (ETDEWEB)

Public Law 95-209 includes a requirement that the Advisory Committee on Reactor Safeguards submit an annual report to Congress on the safety research program of the Nuclear Regulatory Commission. This report presents the results of the ACRS review and evaluation of the NRC safety research program for Fiscal Year 1983. The report contains a number of comments and recommendations.

Wastenet

... Targeted fields of research Continuation of ongoing research - Finalising detailed design work on the ITER project; getting JET operational at full power; Improvement of the basic concepts of fusion devices - Fusion plasmas; theoretical studies; technology watch on research into inertial confinement; new experimental concepts and systems; etc.; Long-term technology - Preparations for building a demonstration reactor (development of tritium breeding blankets; prospective ...

International Nuclear Information System (INIS)

Tarapur Atomic Power Station unit-4 is first 540 MWe pressurized heavy water reactor. Heavy water is used as the coolant and moderator. With reactor operation tritium is formed by absorption of neutron by deuterium atoms. Experience in the 220 MWe indicates that tritium is main contributor to the internal dose of radiation workers. Study on tritium build up in Primary Heat Transport (PHT) and Moderator (MOD) system was carried out at the initial stage of the operation of the unit-4. This paper brings out tritium activity buildup in the PHT and MOD systems and its comparison with 220 MWe reactors. This paper helps in estimation of the internal dose contribution to the radiation workers at TAPS 3 and 4. (author)

Energy Technology Data Exchange (ETDEWEB)

Conference paper regarding research in freeze-protection methods that could extend market acceptance for passive solar domestic water heating systems in more northern climates if the U.S.

International Nuclear Information System (INIS)

In equipment industries, the equipments handling industrial water and pure water are numerous. In power generation including nuclear power generation, water serves as a working medium. Review is made on the experiences in the corrosion of iron and nickel base alloys in high temperature, high pressure water and the results of researches derived from them. Under high temperature and high pressure, carbon steel, low alloy steel, stainless steel and high nickel alloy cause corrosion even in pure water. But in the case of serious corrosion, chlorine, oxygen, alkali and others in water take part. The following matters are described: corrosion by steam; stress corrosion cracking in pure water; corrosion by impurities in high temperature, high pressure pure water, i.e. chlorine ions, dissolved oxygen, and ...

International Nuclear Information System (INIS)

... University, Fracture and Reliability Research Institute, Sendai, Miyagi (Japan)

Energy Technology Data Exchange (ETDEWEB)

This paper describes a wide range of technical research and development that can reduce costs and improve technology for deep water deployment.

International Nuclear Information System (INIS)

The IGC Highlights briefly outlines some of the the significant progresses made by Indira Gandhi Centre for Atomic Research, Kalpakkam during the period 1996-1997. The Fast Breeder Test Reactor (FBTR) was operated at the maximum power level possible with the available partial core. The first generation of electricity from FBTR and its synchronization with the grid in 1997 marked a significant step in the nuclear programme of the Centre. Another important event was the commissioning of the "2"3"3U - fuelled Kamini reactor.The mission-oriented programmes in fast reactor technology was supported by a host of research and development programmes, in closely related areas namely materials technology, welding metallurgy, sodium technology, manufacturing technology, non-destructive testing, quality engineering, in-service inspection, electronics and instrumentation and safety ...

International Nuclear Information System (INIS)

This paper describes an electronic imaging system for producing real time neutron radiography from a low power research reactor, which will allow inspections of samples with high efficiency, in terms of measuring time and result analysis. This system has been implanted because of its potential use in various scientific and industrial areas where neutron radiography with photographic film could not be applied. This real time system is installed in neutron radiography facility of Argonauta nuclear research reactor, at the Instituto de Engenharia Nuclear of the Comissao Nacional de Energia Nuclear, in Brazil. It is adequate to perform real time neutron radiography of static and dynamic events of samples.

Energy Technology Data Exchange (ETDEWEB)

This is the final report of the Finnish Research Programme on Climate Change (SILMU). This report includes the final results and conclusions made by the individual research groups. The aim of this report is to lay out the research work, and to present the main results and conclusions obtained during the six-year work. The Finnish Research Programme on Climate Change (SILMU) was a multidisciplinary national research programme on climate and global change. The principal goals of SILMU were: (1) to increase our knowledge on climate change, its causes, mechanisms and consequences, (2) to strengthen the research on climate change in Finland, (3) to increase the participation of Finnish researchers in international research programmes, and (4) to prepare and disseminate information for policy makers on adaptation and ...

International Nuclear Information System (INIS)

Full text: The transmutation of nuclear waste to reduce the burden on a geological repository is a relevant topic within the Program of Nuclear Safety Research of the Research Centre Karlsruhe. Several studies have confirmed that a high efficiency of transmutation of actinides is reached in fast neutron spectrum reactor system. Therefore, an important effort is dedicated to the study of transmutation strategies with different fast reactors and their associated technologies. Moreover, in international contexts as Generation IV International Forum (GIF) and Sustainable Nuclear Energy Technology Platform (SNETP), fast reactors are considered in the frame of sustainable development of nuclear energy and reduction of waste. The systems that are currently under investigation, in the frame of the different fuel cycle scenarios, are liquid metal cooled and gas cooled fast ...

International Nuclear Information System (INIS)

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized ...

Energy Technology Data Exchange (ETDEWEB)

An inorganic particulate adsorbent of a titania-alumina is described for treating a superheated water containing radioactive materials such as cobalt ions, which is free from release of corrosive impruities, and which has a high adsorption capacity of radioactive materials and a high mechanical strength is prepared by hydrolyzing a titanium alkoxide and an aluminum alkoxide, thereby forming a hydrous titanium oxide and a hydrous aluminum oxide, respectively; precalcining the hydrous titanium oxide and aluminum oxide, mixing and molding the resulting titania and alumina into a particulate mixture thereof having a titania mole fraction of 0.2 to 0.9, and calcining the particulate mixture at 500/sup 0/-700/sup 0/C. This absorbent is effectively used in treat boiling water-type nuclear reactor core-circulating water to remove radioactive substances therefrom.

International Nuclear Information System (INIS)

... 978-5-94883-072-8 121 p. SPECIFIC NUCLEAR REACTORS AND

Energy Technology Data Exchange (ETDEWEB)

As evidence of its effectiveness rapidly accumulates, the Lomi process has most recently been used to decontaminate the recirculation loops and the reactor water clean-up unit of a BWR at Monticello in the United States. An average decontamination factor of 23 was achieved in the recirculation loops.

International Nuclear Information System (INIS)

The 105-KE and 105-KW Fuel Storage Basins were constructed more than 35 years ago as repositories for irradiated fuel from the K East and K West Reactors. Currently, the basins contain irradiated fuel from the N Reactor. To continue to use the basins as desired, seismic adequacy in accordance with current US Department of Energy facility requirements must be demonstrated. The 105-KE and 105-KW Basins are reinforced concrete, belowground reservoirs with a 16-ft water depth. The entire water retention boundary, which currently includes a portion of the adjacent reactor buildings, must be qualified for the Hanford Site design basis earthquake. The reactor building interface joints are sealed against leakage with rubber water stops. Demonstration of the seismic adequacy of these interface joints was initially identified as a key issue in the ...

Science.gov (United States)

Liquid Phase Catalytic Exchange (LPCE)- Cryogenic Distillation(CD) process was studied which could be available for an optimal tritium removal process of pressurized heavy water reactor system at Wolsung nuclear power plant in the near future. Based upon ...

Science.gov (United States)

The Wolsung-type nuclear fuel localization project carried out since 1981 finally reached to a full-fledged phase in 1987. We successfully produced and timely delivered a yearly demand of nuclear fuel for Wolsung unit 1. In this report we studied and summ...

International Nuclear Information System (INIS)

Full text: There are following general tendencies linking to SNF and radioactive waste management (RWM) in the Russian nuclear industry now. The intention to use the closed nuclear fuel cycle based on power water reactors and fast reactor. The intensification of measures aimed at the solution of 'nuclear legacy' from defenses programs of USSR. The intention to improve the existing national RW management infrastructure in the near years by means of the creation of a centralized national system (including managing corporation responsible for operation of long-storage and disposal facilities of conditioned RW). The main aims radioactive waste management (RWM) in nuclear power plants (NPP) for the next 10-15 years are to equip all NPPs with the necessary set of facilities for conditioning of the stored and currently generated RW with packaging the end-product into containers, to build regional NPPs RW repositories and to ...

International Nuclear Information System (INIS)

In the present paper, experimental study of forced convection boiling were performed to investigate the post-CHF characteristics of a vertical annular channel with one heated rod and four spacer grids for new refrigerant R-134a. The experiments were conducted under outlet pressure of 11.6, 13, 16 and 20 bar, mass fluxes of 100-600 kg/m2s, and inlet temperatures of 25-51 .deg. C. The parametric trend of the post-CHF data was well consistent with previous studies. The two phase flow regime in tube flow occurring downstream of the CHF has been called post-CHF, dispersed flow, liquid-deficient flow, mist flow and film boiling. This regime is characterized by a continuous vapor phase with discrete liquid drops and a non-wetted heated surface. This regime has a considerable importance in the areas of light water reactor(LWR) accident analysis and other film boiling applications. The post-CHF region occurs by design in heat exchangers operating in the ...

Energy Technology Data Exchange (ETDEWEB)

This report is a summary of the contribution of nitrogen and phosphorus in the eutrophication process of inland and coastal waters. Special attention was paid to the mechanisms of these nutrients in regulating biological processes and to the methods available in estimating their effects in the eutrophication of water bodies. The report includes five chapters which are entitled: Introduction, which is a general background to the subject with special attention to the requirements of the Finnish Water Act. Phosphorus and nitrogen as factors regulating biological processes. The topics included are: definition of eutrophication, forms of phosphorus and nitrogen and their sources to inland and coastal waters, effects of these nutrients as growth factors of phytoplankton and macrophytes and consequences of eutrophication. Estimation of the effects of phosphorus and nitrogen. The topics discussed from the point ...

Energy Technology Data Exchange (ETDEWEB)

Progress is briefly summarized for a large number of projects in the areas of research instruments, measurement and controls engineering, reactor systems, and maintenance management. (LEW)

International Nuclear Information System (INIS)

An important application of metal hydrides is as a moderator material in nuclear reactors. The fundamental properties of hydrides are illustrated and an impression given of the current research into hydrogen in transition metals. Phase diagrams, magnetic properties, temperature dependence of the diffusion coefficient, energy level schemes and superconductivity are considered. (C.F.).

International Nuclear Information System (INIS)

A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization.

Energy Technology Data Exchange (ETDEWEB)

The activation aspects of pure fusion and hybrid fusion technology is studied to assess the radioactive safety of various fusion concepts including tokamak pure fusion, fissile fuel producing hybrid and radio waste transmuting hybrid. The activation properties of breeding, coolant and structural materials in fusion reactors might be quite different from those in fission reactors because of the high energy D-T fusion neutrons from the fusion cores. A study on the involved activation reactions and the uncertainties of the associated nuclear cross-sections is carried. The activation properties of various first wall concepts and blanket concepts are discussed. The radioactive inventory during the operation lifetime and the potential hazard of the radioactive nuclides with respect to near term (reprocessing) and long term (waste disposal) aspects are calculated, with reference to ITER/NET (International Thermonuclear Experiment ...

International Nuclear Information System (INIS)

The radiation dose and the atmospheric dispersion for Pressurized Heavy Water Reactors (PHWR) are based on the CAN/CSA N288.2-M91 standards: for Pressurized Water Reactor (PWR) on the NRC Regulatory Guide 1.145. There are some differences between in the methodologies used in the standards, including the atmospheric dispersion model, the release height, the temperature lapse rate, the cutoff condition. This paper reports on a comparison of standards for atmospheric dispersion models of PHWRs and PWRs in order to determine which one is the more conservative. The comparison between PHWR and PWR for atmospheric dispersion factors and radiation doses confirms that there are no big differences

Energy Technology Data Exchange (ETDEWEB)

The results of an investigation of the effects of several parameters on the reactivity instability of a Boiling Water Reactor (BWR) calculational model are summarized. Calculations were performed for a typical BWR operated at low flow conditions, where reactivity instabilities are more likely to occur. The parameters investigated include the axial power shape (characterized by two separate parameters), the core pressure, and operating flow. All calculations were performed using the LAPUR code which was developed at the Oak Ridge National Laboratory for the dynamic modeling of large BWR's. 4 refs., 8 figs.

International Nuclear Information System (INIS)

The temperature coefficient has been investigated on the Wolsung nuclear power reactor, in which fuel is natural uranium dioxide and moderator heavy water. The numerical computations are carried out in terms of changes of the effective neutron multiplication factor with respect to fuel, moderator, and coolant temperatures. Those results are compared with the computed values of temperature coefficient based on the LATREP computer code. (author).

International Nuclear Information System (INIS)

As the cost of developing completely new computer codes becomes prohibitive, designers of nuclear facilities are turning to more cost-effective approaches for meeting increasingly strict regulatory requirements applied to safety-related analysis. For designing and licensing the MAPLE family of research reactors, Atomic Energy of Canada Ltd. (AECL) is employing the strategy of adapting major existing codes by linking them together within networks of custom-built interface software. This approach builds on the international investment in developing, maintaining, and verifying existing primary codes and focuses on the less onerous development of interface codes. The resultant code systems are then validated for the new applications of interest.

British Library Electronic Table of Contents (United Kingdom)

Determination of thermal to fast neutron flux ratio (ffast) and fast neutron flux (phi-fast) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The ffast and subsequently phi-fast were determined using the absolute method. The ffast ranged from 48 to 155, and the phi-fast was found in the range 1.03x1010-4.89x1010 n cm-2 s-1. These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

International Nuclear Information System (INIS)

The characteristics of real-time neutron radiography are described briefly in this paper, and the acquirement of neutron flux, the selection of convertor and the structure of the twilight imaging system and the image-sampling and image-processing system in SPRR-300 reactor are also analyzed detailedly. The experimental result of real-time neutron radiograph is too analyzed in this paper

International Nuclear Information System (INIS)

An AHR (Advanced HANARO Reactor) based on the HANARO has been conceptually developed for the future needs of research reactors. Generally, a natural convection cooling in nuclear installations is an ultimate heat removal mechanism as an inherent safety feature. This paper presents the preliminary thermal hydraulic characteristics and safety margins for a natural convection cooling in the AHR.

Science.gov (United States)

Hydrothermal oxidation (HTO) is a promising technology for the treatment of aqueous-fluid hazardous and mixed waste streams. Waste streams identified as likely candidates for treatment by this technology are primarily aqueous fluids containing hazardous organic compounds, and often containing inorganic compounds including radioisotopes (mixed wastes). These wastes are difficult and expensive to treat by conventional technologies (e.g. incineration) due to their high water content; in addition, incineration can lead to concerns related to stack releases. An especially attractive potential advantage of HTO over conventional treatment methods is the total containment of all reaction products within the overall system. The potential application of hydrothermal oxidation (HTO) technology for the treatment of DOE hazardous or mixed wastes has been uncertain due to concerns about safe and efficient operation of the technology. In principle, aqueous DOE wastes, including ...

Wastenet

... Boreal Forest and Climate Change Pertti Hari, Liisa Kulmala, 160.45 EUR; 209.00 USD; 135.00 GBP... Global Atmospheric Change and its Impact on... Ian Barnes 145.00 USD; 128.35 EUR; 95.00 GBP... More Journals » MAGAZINES & JOURNALS Journal of Water Supply: Research &Technology -... Journal of Water Supply: Research and Technolog... ...

Energy Technology Data Exchange (ETDEWEB)

The Western Research Institute plan for addressing oil shale plant siting methodology calls for identifying the available resources such as oil shale, water, topography and transportation, and human resources. Restrictions on development are addressed: land ownership, land use, water rights, environment, socioeconomics, culture, health and safety, and other institutional restrictions. Descriptions of the technologies for development of oil shale resources are included. The impacts of oil shale development on the environment, socioeconomic structure, water availability, and other conditions are discussed. Finally, the Western Research Institute plan proposes to integrate these topics to develop a flow chart for oil shale plant siting. Western Research Institute has (1) identified relative topics for shale oil plant siting, (2) surveyed both published and ...

Energy Technology Data Exchange (ETDEWEB)

Four 10 MW, highly-stabilized power supply modules have been installed at the National High Magnetic Field Laboratory in Tallahassee, FL, to energize water-cooled, resistive, high-field research magnets. The power supply modules achieve a long term current stability if 10 ppM over a 12 h period with a short term ripple and noise variation of <10 ppM over a time period of one cycle. The power supply modules can operate independently, feeding four separate magnets, or two, three or four modules can operate in parallel. Each power supply module consists of a 12.5 kV vacuum circuit breaker, two three-winding, step-down transformers, a 24-pulse rectifier with interphase reactors, and a passive and an active filter. Two different transformer tap settings allow rated dc supply output voltages of 400 and 500 V. The rated current of a supply module is 17 kA and each supply module has a one-hour overload capability of 20 kA. The ...

International Nuclear Information System (INIS)

Calculations have been made to determine the production of "1"8"8W from "1"8"6W in several US fission reactor systems, e.g., Fast Flux Test Facility (FFTF), the High Flux Isotope Reactor (HFIR), and the Advanced Test Reactor (ATR). Important input to these calculations are the cross-section parameters for "1"8"6W, "1"8"7W, and "1"8"8W. Only two values have been measured for "1"8"7W and none for "1"8"8W. Consequently, results from integral measurements play a crucial role in determining the "1"8"7W and "1"8"8W values. This has been studied for irradiations in the FFTF and the Oregon State Univ. (OSU) research reactor. Short irradiation of enriched "1"8"6W in both the FFTF and the OSU reactors have produced #mu#Ci/g quantities of "1"8"8W/"1"8"8Re. Measurements were made of the "1"8"8W gamma ray emission. These results were incorporated with other available data to ...

International Nuclear Information System (INIS)

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. The defect rate was twice that in 1978 but still lower than the two previous years. Methods being employed to detect defects include increasing use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failures by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. (author).

International Nuclear Information System (INIS)

The Fast Flux Test Facility has been in operation for approximately five years, including about one thousand days of full power operation of the Fast Test Reactor. During that time the collective dose equivalents received by operating personnel have been about two orders of magnitude lower than those typically received at commercial light water reactors. No major contamination problems have been encountered in operating and maintaining the plant, and release of radioactive gas to the environment has been minimal and well below acceptable limits. All shields have performed satisfactorily. Experience to date indicates an apparent radiological superiority of liquid metal reactor systems over current light water plants.

Energy Technology Data Exchange (ETDEWEB)

At present, two types of reactors, Pressurized Light Water Reactor(PLWR) and Pressurized Heavy Water Reactor(PHWR), are operating and under construction in Korea. They are much different in design concepts and inherent features from each other so that the calculation methods for Exclusion Area Boundary(EAB) are also different from each other. Thus, the domestic calculation methodology has been applied to PHWR, Wolsung 2, 3 and 4. In this report, the regulatory requirements and methodologies for EAB of Canadian methodology for EAB has been also investigated. It has been examined that the Canadian methodology which has been applied to the calculation of EAB of Wolsung 2, 3 and 4 can be said to be conservative enough compared to physical phenomena. 4 tabs., 3 figs., 22 refs. (Author).

Energy Technology Data Exchange (ETDEWEB)

Intelligent and decision aiding systems as support to operators are becoming increasingly a necessity in nuclear installations and in nuclear reactors in particular, specially after the Tree Mile Island. Development of new technologies based on linguistic approaches such as fuzzy logic has given rise to much interest during the last years. Fuzzy logic controller (FLC) has many advantage compared to conventional controllers using classical techniques. The aim of the present work is to use a fuzzy logic controller in parallel to actual semi-automatic controller in order to supervise in real time the operation of the research nuclear reactor. The principal of this controller is based on rules which are established previous from experiment using the semi-automatic controller and from the knowledge of the operators. (authors)

British Library Electronic Table of Contents (United Kingdom)

Water?rock interaction is one of the prime factors affecting the fluoride contents of surface and groundwater. If fluoride concentration of drinking water has been neglected, excess fluoride can cause serious dental and medical problems on human health, which is well known at Golcuk-Isparta region. In the research area, Egirdir lake, Golcuk lake and surrounding springs have been utilized as drinking water sources. Golcuk lake water and surrounding groundwaters have high fluoride content (1.4?4.6?mg/l), which is above the WHO standards. Fluoride is predominantly supplied by dissolution of fluoride within the fluormicas of volcanics during the circulation of water. Fluoride concentrations of waters have shown variations for dry and rainy seasons depending on the degree of interaction between...

Energy Technology Data Exchange (ETDEWEB)

This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

Energy Technology Data Exchange (ETDEWEB)

The Nova Scotia Research Foundation Corporation was established in 1946 by the Province of Nova Scotia to use science and technology to assist in the economic development of Nova Scotia. Within the Applied Science Division of the Corporation, research and development programs in the area of coal and mineral technology have been undertaken. A brief description of two projects -- coal liquefaction and coal-water fuels, are reported.

Energy Technology Data Exchange (ETDEWEB)

Previous experiments in fast critical assemblies demonstrated a method of determining reactivity from power spectral density measurements with /sup 252/Cf. This method determines reactivity from properties of the reactor only at the subcritical state of interest, thus it does not require a calibration near delayed criticality. The interpretation of the measured data to obtain reactivity does not require knowledge of the relative or absolute values of the source intensity, knowledge of the detection efficiencies, or knowledge of the detection instrumentation frequency responses. An experiment was performed at the Pool Critical Assembly to evaluate the possibility of /sup 252/Cf source driven neutron noise spectral density measurements in light water moderated reactors. This experiment showed that using commercially available detectors, such measurements can be performed in a reasonable time, that is, the measurement of the ...

International Nuclear Information System (INIS)

In the radiochemical reprocessing of spent fuel from nuclear reactors, a considerable amount of liquid, solid, and gaseous waste is formed; this waste is potentially dangerous to humans and requires the development of special and complex technological techniques for its localization and reliable long-term storage. The most hazardous are liquid wastes of high specific activity - water-tailings solutions obtained in the first cycle of extraction after the removal of uranium and plutonium. These solutions contain more than 99.9% of all the other transuranic elements - isotopes of neptunium, americium, and curium. Where necessary, some fission products and actinides may be removed from wastes of high specific activity for subsequent use. The quantity, composition, and activity of these wastes varies within broad limits, depending on the type and power of the reactor, the initial nuclide composition of the fuel, and its specific ...

International Nuclear Information System (INIS)

The evaluation of the hazard posed to the skin by very small radioactive sources (diameter < 1mm) has become popularly known as the 'hot particle' problem in European and American nuclear reactor facilities. In this study, research to detect hot particle was performed in Wolsung Nuclear power plant (NPP) in Korea.

Energy Technology Data Exchange (ETDEWEB)

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled ...

Energy Technology Data Exchange (ETDEWEB)

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs.

Energy Technology Data Exchange (ETDEWEB)

The ORR operated at an average power level of 29.7 MW for 85.3% of the time during this period. The reactor was shut down on fifteen occasions, nine of which were unscheduled. Reactor downtime needed for refueling and checks was normal. The reactor remained available for operation 88.3% of the time. Special tests completed during the quarter included: (1) transfer of LEU fuel elements CLE-202 and NLE-201 from core positions B-9 and B-2 to core positions C-5 and C-6 for continued operation; and (2) calculation of maximum heat flux in LEU elements CLE-201 and NLE-202 in core positions A-2 and A-8. In-service inspections included inspections of ORR decay tank, primary heat exchanger No. 4, and the 24-in. strainer.

International Nuclear Information System (INIS)

The paper presents the progress of the Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor WWR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for WWR-S decommissioning was also developed. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is ...

Energy Technology Data Exchange (ETDEWEB)

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to house the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for ...

International Nuclear Information System (INIS)

The Gross Decontamination Experiment was conducted on various levels and surfaces of the TMI-2 Reactor Building during February and March 1982 and was designed to investigate the effectiveness of various surface decontamination techniques. The polar crane, D-rings, missile shields, refueling canal, fueling bridge, major equipment, floors and some walls were flushed with low pressure water. Water lances were directed manually and applied water at temperatures between ambient and 60"0C at a flow rate of about 95 liters per minute. In addition, floor surfaces on the 305-ft elevation and floor surfaces and major equipment on the 347-ft elevation were sprayed with high pressure water (floors in the Reactor Building are designated by their elevations above sea level). The water pressure in this case varied between 13.8 and 41.4 mPa and ...

International Nuclear Information System (INIS)

Inherent boron dilution can occur in case of a Small Break LOCA when low borated water is mainly accumulated in the U-legs due to reflux boiling in the Steam Generator tubes after the loss of natural circulation. The restart of the natural circulation may lead to criticality because of the injection of these low borated slugs towards the core. To evaluate this potential risk, the boron concentration at the core inlet has to be known which makes necessary to estimate the mixing phenomena in the cold leg, in the downcomer and in the lower plenum: CFD calculations are required. First of all the validation of CFX5 CFD code on the relevant phenomena of inherent boron dilution has been established (UPTF TRAM C3 test). Then, an application to the 900 MW French Pressurized Water Reactor series has been performed. (authors)

International Nuclear Information System (INIS)

The University of Michigan (UM) is a member of a team of researchers, including the universities of Florida, Texas, and Tennessee, along with Oak Ridge National Laboratory, developing robotic for hazardous environments. The goal of this research is to develop the intelligent and capable robots which can perform useful functions in the new generation of nuclear reactors currently under development. By augmenting human capabilities through remote robotics, increased safety, functionality, and reliability can be achieved. In accordance with the established lines of research responsibilities, our primary efforts during 1991 will continue to focus on the following areas: radiation imaging; mobile robot navigation; three-dimensional vision capabilities for navigation; and machine-intelligence. This report discuss work that has been and will be done in these areas.

British Library Electronic Table of Contents (United Kingdom)

This study deals with the neutronics analysis of the current core configuration of a 3MW TRIGA Mark-II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Safety Analysis Report (SAR) values. The comprehensive neutronics code system SRAC was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Cross-section data library generated from JENDL-3.2 were used. The validation of the model against benchmark experimental results is presented. The SRA...

Energy Technology Data Exchange (ETDEWEB)

During FY1997, the JMTR was operated for 3 complete cycles (120th, 121st and 122nd cycles) and was utilized for the research and development programs on the technology of LWRs and fusion reactor, as well as for fundamental research of fuels and materials, and for radioisotope productions. The improvement of evaluation technique in a local neutron spectrum for irradiation utilization and development of capsule having the vertical migration, the reinstrumentation and loading mechanism have been carried out. Development of a new oxygen potential sensor for oxide fuel pellets has been done as an elemental technology of irradiation for high burn-up fuels. As for post irradiation examination, the techniques for measuring of crack length using an alternating current potential drop method and machining of miniaturized specimen by the remote handling have been developed. A research on the blanket materials and ...

Energy Technology Data Exchange (ETDEWEB)

Nuclear energy related research in Switzerland is concentrated at PSI's Nuclear Energy and Safety Research Department (NES). The activities of the department are concentrated on three main domains of: Safety and related problems of operating plants; safety features of future reactor and fuel cycles; waste management. Comprehensive assessments of energy systems are carried out in cooperation with PSI's General Energy Research Department. Many of the programs are part of collaborations with universities, industry, or international organisations. Progress in 2000 in these topical areas is described in this report. A list of scientific publications in 2000 is also provided.

Energy Technology Data Exchange (ETDEWEB)

This report, the eighth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1992. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that a number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with the technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the ...

International Nuclear Information System (INIS)

Full text: In heavy-water-moderated power reactors, tritium is primarily produced by neutron capture in deuterium nuclei in the moderator and coolant. For CANDU 6 reactors, the estimated steady-state values are #approx# 3 TBq#centre dot#kg"-"1 D_2O in the moderator and #approx# 74 GBq#centre dot#kg"-"1 D_2O in the coolant. Tritium removal is one option available to reactor operators for use in their heavy water and tritium management strategies. The WTRF is designed to remove tritium from tritiated heavy water in each of the four CANDU units at the Wolsung Site, to immobilize the tritium and to store it on site. The detritiation process is based on three steps: the first one (front-end) involves the transfer of tritium from heavy water to deuterium gas; the second one (enrichment) concentrates the tritium in a cryogenic distillation system ...

International Nuclear Information System (INIS)

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown ...

International Nuclear Information System (INIS)

... power reactors pressure tube reactors reactors THERMAL REACTORS.

Science.gov (United States)

The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the ...

International Nuclear Information System (INIS)

Taking into consideration India's limited reserves of natural and vast reserves of thorium, the fast reactor route holds a great promise for India's energy supply in future. The fast reactor fueled with "2"3"9Pu/"2"3"8U (unused or depleted) produces (breeds) more fissionable fuel material "2"3"9Pu than it consumes. Calculations show that a fast breeder reactor (FBR) increases energy potential of natural uranium by about 60 times. As the fast reactor can also convert "2"3"2Th into "2"3"3U which is a fissionable material, it can make India's thorium reserves a source of almost inexhaustible energy supply for a long time to come. Significant advantage of FBR plants cooled by sodium and their world-wide operating experience are reviewed. There are two main safety issues of FBR, one nuclear and the other non-nuclear. The nuclear issue concerns core disruptive accident and the non-nuclear one concerns the ...

Energy Technology Data Exchange (ETDEWEB)

This report summarizes the major activities conducted in the Chemical and Energy Research Section of the Chemical Technology Division (CTD) at Oak Ridge National Laboratory (ORNL) during the period January--March 1997. Created in March 1997 when the CTD Chemical Development and Energy Research sections were combined, the Chemical and Energy Research Section conducts basic and applied research and development in chemical engineering, applied chemistry, and bioprocessing, with an emphasis on energy-driven technologies and advanced chemical separations for nuclear and waste applications. The report describes the various tasks performed within seven major areas of research: Hot Cell Operations, Process Chemistry and Thermodynamics, Molten Salt Reactor Experiment (MSRE) Remediation Studies, Chemistry Research, Separations and Materials Synthesis, ...

International Nuclear Information System (INIS)

... brakes lmfbr type reactors breeder reactors epithermal reactors fast reactors

International Nuclear Information System (INIS)

Water chemistry of nuclear power plants has played an important role in reduction of personnel doses, structural materials and fuel integrity assurance, and reduction of radioactive wastes production. Further contributions are requested for advanced utilization of the LWR, advanced fuels and aging management of plants. Since water chemistry has an effect on all structure and materials immersed and at the same time affected by them, the optimum control not sticking to specific issues and covering the whole plant is required for these requests. Taking account of roles and activities of the industry, governmental institutes and academia, road maps on research and development plans for water chemistry were compiled into identified eleven items with targets and counter measures taken, such as common basic technologies, dose reduction, SCC mitigation, fuel cans corrosion/hydrogen absorption mitigation, ...

Energy Technology Data Exchange (ETDEWEB)

The Coastal Zone Sector of the Canadian Climate Impact and Adaptation Research Network (C-CIARN) was established to address concerns regarding climate change in coastal areas. Coastal zones are sensitive to increases in air, sea and ground temperatures as well as to variations in sea level, precipitation, ice thickness, and storm intensity. This report presents the results of a workshop which focused on how coastal communities will be affected by climate induced water level changes, particularly sea-level rise on ocean coasts and decreasing lake levels in the Great Lakes. The workshop addressed issues such as the effects of changing water levels on coastal infrastructure, utilities, community development, and the implications of changing water levels to human safety, cultural resources, tourism, and insurance. The workshop reviewed the current state of understanding on water level ...

Energy Technology Data Exchange (ETDEWEB)

Condensing two-phase channel flow occurs in many industrial applications, including heating and refrigeration systems. It can also occur in certain nuclear reactor accidents. For example, during a small-break loss-of-coolant accident in a pressurized water reactor, following the partial depletion of the primary coolant, condensation of steam on the primary side of the steam generator tubes can provide a heat sink for disposal of the decay heat generated in the reactor core. Condensing two-phase flow can also play an important role in the operation of the passive emergency cooling system in the advanced simplified boiling water reactor. Here, steady-state condensation in the presence of a noncondensable in a concurrent two-phase channel flow is analyzed using a two-fluid model. The effect of noncondensables on the combined heat transfer at the liquid-gas mixture ...