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)

This paper discusses the remote plugging of leaks in inaccessible pipework, with main reference to small leaks which frequently appear in ancillary cooling water circuits of nuclear reactors. Initially developed to cure problems of the pre-stressed concrete pressure vessels of UK reactors, the ZORIC sealant has been used to repair leaking biological shield pipework on six CANDU reactors. ZORIC is based on a water-soluble epoxy resin and an aqueous suspension of a refined mineral clay. This paper describes the evolution of the sealant, the qualification and testing program, and their application to CANDU reactor systems. 2 refs., 6 figs.

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 ...

Energy Technology Data Exchange (ETDEWEB)

This paper presents the objective, scope, and status of the US Department of Energy's Materials and Structures Program to develop a data base on mechanical properties of structural materials for out-of-core structures and components for LMFBRs. Information on the development of a reference data base on materials for the reactor system, reactor enclosure system, primary heat transport system, intermediate heat transport system, and steam generator system is included. In addition, the development of the data and analyses to account for the effects of temperature and stress, as well as water/steam, sodium, and radiation environments, is described. Plans for the development of alternative materials for future out-of-core applications are presented.

International Nuclear Information System (INIS)

RELAP5 (Reactor Excursion and Leak Analysis Program) is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor UMLRR are used. The UMLRR is a 1 MW, light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the development and the validation of a UMLRR-RELAP model using experimental data. For this ...

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)

In order to realize improve of reliability and economy by duplicate production, rapid supply of repair parts from standardized storage, such were expected as to have continuous order of standardized plant, to ignore site condition, to avoid expansion of regulatory requirement. Standardization program was planned to limitedly promote standardization of safety-related design concept, major specification and basic system composition of reactor and primary systems. The area of standardization had been tried to expand to BOP such as general arrangement and rad-waste system.

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 ...

International Nuclear Information System (INIS)

A program under the sponsorship of the United States Nuclear Regulatory Commission was intiated at the Oak Ridge National Laboratory (ORNL) in late 1977. The program, Advanced Instrumentation for Reflood Studies (AIRS), is charged with developing instrumentation for measurement of in-vessel fluid phenomena in pressurized water reactor reflood facilities. The goal of the ORNL program is to develop techniques and systems for measuring fluid flow in-core, deentrainment in the upper plenum and liquid fallback from the upper plenum into the core. A large portion of the development at ORNL is devoted to the impedance probes for measurement of two-phase flow velocities and void fractions. Film probe development at ORNL is limited to adapting the present techniques to the environment of a reflood facility. As the development progresses on all the measurement techniques, ORNL will fabricate ...

Energy Technology Data Exchange (ETDEWEB)

The objective of Configuration Risk Management Program(CRMP) is to maintain the safety level by assuring the defense-in-depth of nuclear power plant while the configurations are changed during plant operations, especially for the LP/SD. Such a safety purpose can be achieved by establishing the risk monitoring programs with both quantitative and qualitative features. Generally, the quantitative risk evaluation models, i.e., PRA models are used for the risk evaluation during full power operation, and the qualitative risk evaluation models such as safety function assessment trees are used. Through this study, safety function assessment trees were developed.

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 ...

Science.gov (United States)

The Heavy Water Reactor Facility (HWRF), as part of the defense-in-depth philosophy to mitigate the effect of design-basis and severe accidents, is equipped with a passive containment cooling system (PCCS). The function of the PCCS is to provide a safety-grade path to the ultimate heat sink for the removal of the reactor coolant system sensible heat and core decay heat. Ambient air enters an annular space between the steel containment shell and the surrounding concrete shield building through inlets in the shield building wall, is heated via natural convection, rises, and exits the building through a chimney located above the containment dome. A test program is in place to access parameters important to the effective operation of the PCCS. This paper focuses on the large-scale tests (LSTs). The objectives of these tests are as follows: (1) demonstrate natural circulation cooling with more prototypic ...

International Nuclear Information System (INIS)

Ppercase(femaxi-iv) is a computer code for the analysis of thermal and mechanical behavior of light water reactor fuel rods during steady-state and transient conditions. The main purpose of ppercase(femaxi-iv) is to calculate the stress and strain distributions in the fuel and cladding due to the pellet-cladding mechanical interaction, and the fission gas release rate during operations, especially power transients. The capability of ppercase(femaxi-iv) was extensively tested with a number of experimental results obtained in some international fuel irradiation programs. This paper provides a general description of the various models involved in ppercase(femaxi-iv) and some results of applications on the experimental data. ((orig.)).

Energy Technology Data Exchange (ETDEWEB)

The costs of establishing a computer-based training program for FFTF reactor operators are analyzed.

Energy Technology Data Exchange (ETDEWEB)

The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event.In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping.In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were ...

Science.gov (United States)

Computer programs have been developed to define the temperature increase which would be needed to bring deep-ocean water into density equilibrium with surface water for locations where data are available. A series of continuous-flow studies on phytoplankton blooms resulting from mixtures of 80 percent deep and 20 percent surface water in 2000-liter concrete culturing vessels (''reactors'') has been completed. A quantitative determination of nutrient utilization and flow through a combined primary and secondary trophic level system has been completed. This study utilized the clam Tapes semidecussata, fed from phytoplankton grown in 80 percent deep and 20 percent surface water. An analysis of the fate of the deep water discharged from a floating OTEC plant indicates that horizontal containment of the ...

International Nuclear Information System (INIS)

Argentina has two operating PHWR nuclear power plants. Atucha I NPP is a pressure vessel type heavy water reactor of 360 MW e with 25 years of operation and Embalse NPP is a pressure tube type CANDU-600 reactor of 640 MW e. Atucha II, a third plant of 600 MW e of the pressure vessel type similar to Atucha I, is being constructed. NASA (Nucleoelectrica Argentina S.A.) currently operates both nuclear power plants. The National Atomic Energy Commission (Comision Nacional de Energia Atomica - CNEA) provides operational support to the plants, including research and development assistance, and actual technical services and maintenance work in different areas. The Chemistry Department, formerly the Reactor Chemistry Department has carried out project and support activities to the plants during the past 20 years. The aim of this work is to describe the present organization and the activities in ...

International Nuclear Information System (INIS)

The CERL Code was developed to describe the solution chemistry of the water on the steam generating side of PWR reactors. It is designed to calculate the equilibrium species distribution resulting from the interaction of impurities, corrosion products, and additives in the aqueous solution. It calculates the extent of ion-ion interactions, the precipitation of insoluble species and the amount of solute that partitions into the vapor phase when some of the water evaporates. This knowledge of the bulk phase equilibrium distribution of species, especially the pH should be useful in describing the corrosion processes at the solid liquid boundary. The code does not calculate any changes in oxidation states or any rates of reaction. Therefore, it is incapable of calculating the actual corrosion rates. It is anticipated that it will be used as a subprogram of a larger program that will include the redox ...

Science.gov (United States)

An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs. (DG)

International Nuclear Information System (INIS)

In Japan, the development of next-generation Light Water Reactor has been launched since April 2008. The development program will be completed in 2015. The purpose of development is to cope with the replacement for existing nuclear power plants after 2030 in Japan and the expanding demand for nuclear power in the world; 'Nuclear Renaissance.' The reactor also aims to be global standard at around 2030. The requirements for global standard and domestic users have been investigated through the feasibility study of past 2 years, 2006-2007, and six innovative features or 'Core-Concepts' were established as follows. A) Reactor core system with uranium enrichment above 5% for significant decrease of spent fuel discharge and prominent higher availability B) Long-life materials and innovative water chemistry technologies for 80 years plant lifetime and significant ...

KoreaScience (Korea)

Full Text Available

International Nuclear Information System (INIS)

... 1535-DR(Vol.4). Available from INIS in electronic form breeder reactors

Energy Technology Data Exchange (ETDEWEB)

Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously ...

Energy Technology Data Exchange (ETDEWEB)

Recommendations of the Advisory Committee on Reactor Safeguards are presented to the Commissioners for their consideration for FY 82 budget for the NRC safety research program.

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.

International Nuclear Information System (INIS)

This paper discusses how the U.S. Liquid Metal Reactor Development Program has been restructured to carry out R and D on advanced reactor technology. The program gives particular emphasis to improvements to reactor safety. The new directions are based on the technology of the integral fast reactor (IFR). Much of the basis for superior safety performance using IFR technology has been experimentally verified and aggressive programs continue in EBR-II and TREAT. Progress has been made in demonstrating both the metallic fuel and the new electrochemical processes of the IFR. The FFTF facility is converting to metallic fuel; however, FFTF also maintains a considerable U.S. program in oxide fuels. In addition, generic programs are continuing in steam generator testing, materials development, and with ...

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)

Object: To limit the discharge amount of reactor water in a primary system at the time of scram to prevent excessive outflow of reactor water outside the system. Structure: A signal from an upper limit position indicator detects the fact that control rods are completely inserted when the reactor is urgently stopped and the detection signal causes a valve in an outflow line of the discharge water from a control rod driving mechanism to be closed to limit the amount of discharge flown into the scram discharge vessel, thus preventing outflow of reactor water in the primary system after the scram has been initiated. (Kamimura, M.).

International Nuclear Information System (INIS)

Purpose: To stably control the reactor water level so as not to cause excess water feeding in a BWR type reactor. Constitution: A flow control valve is disposed to the exit of a feedwater pump for a nuclear reactor and the valve is controlled by a flow regulator to maintain the water level constant in the reactor. A signal from a water level controller is inputted to the flow regulator to thereby control the flow rate control valve. In this case, the flow regulator remains in a saturated state just after the starting of the feedwater pump, in which the pump flowrate is at 100% to result in an excess water feeding condition. In view of the above, a feedback circuit is provided to the flow regulator so that the saturated state is eliminated and the water feeding can be controlled ...

International Nuclear Information System (INIS)

The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report is volume 1 of the ...

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.

Energy Technology Data Exchange (ETDEWEB)

This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure ...

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 ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

The most important and difficult part of materials research for fusion reactor is realized to be irradiation studies of fusion reactor materials. Irradiation studies of fusion reactor materials utilizing FFTF/MOTA, as one of Japan/U.S.A. Fusion Collaboration Programs, have important role to establish fundamental understanding of heavy irradiation effects on materials behavior and properties and to develop methods and technologies for advanced irradiation studies under fusion reactor environment. This paper briefly reviews the history, the state of the art, and the future of the FFTF/MOTA program. (author).

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.

Science.gov (United States)

A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA).

International Nuclear Information System (INIS)

The status of neutron activation cross sections for some threshold reactions important for reactor materials dosimetry is reviewed. An attempt is made to understand and explain discrepancies between integral and differential data, using recent available experimental results. The importance of standard and benchmark neutron fields for testing differential data for reactor dosimetry is emphasized and the Interlaboratory Reaction Rate (ILRR) program, as well as a similar program pursued by the IAEA, are briefly described.

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 sensitivity/uncertainty analysis methods ...

International Nuclear Information System (INIS)

BARC has organized an international round robin analysis program to carry out the ultimate load capacity assessment of BARC containment (BARCOM) test model. The test model located in BARC facilities Tarapur, is a 1:4 scale representation of 540 MWe pressurized heavy water reactor (PHWR) pre-stressed concrete inner containment structure of Tarapur Atomic Power Station (TAPS) unit 3 and 4. The features of the BARCOM test model and the constitutive data for the pre-test analysis along with the comparison of the results submitted by various participants are described in this pre-test report of the round robin analysis of BARCOM test model

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 ...

Energy Technology Data Exchange (ETDEWEB)

In this paper the possibility of configuring a water cooled Nuclear Thermal Propulsion (NTP) rocket, based on a Particle Bed Reactor (PBR) is investigated. This rocket will be used to operate on water obtained from near earth objects. The conclusions reached in this paper indicate that it is possible to configure a PBR based NTP rocket to operate on water and meet the mission requirements envisioned for it. No insurmountable technology issues have been identified.

International Nuclear Information System (INIS)

Site studies for CANDU type NPP began in a careful manner since 1982 as a first part of the Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. A team was charged to develop all packages of the necessary main studies. The first Romanian NPP CANDU 6 type reactor gone to erection on Cernavoda site, planned to have 5 units and, like Wolsong NPP, applied the same design for the nuclear island. For the BOP parts the ANSALDO-GE project was applied with a thorough concern about requirements raised by connection to NSP. The first mission of design and research multi-branch team was to adapt the NPP Cernavoda project having an open water cooling circuit 'once-through' to the new parameters of a close recirculation water cooling circuit. Also, the structural design was re-evaluated for the case of soft foundation strata instead of hard rock ones. The close ...

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 ...

Energy Technology Data Exchange (ETDEWEB)

Tritium dose management is an important aspect of the radiation protection program at CANDU type reactor sites. This paper describes the bioassay and dosimetry of tritium at CANDU reactor sites, especially for Wolsung Nuclear Power Plant. It presents a compilation of information drawn from published papers, technical reports, international and national guidelines as well as practical experience both in Korean and Canadian CANDU Nuclear Power Plants. The implementation of this program would provide a technical basis for calculations and records should be of acceptable quality and should meet overall radiation protection program objectives.

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. An analysis procedure to model the pipe ...

Energy Technology Data Exchange (ETDEWEB)

Thermal-hydraulics analysis can be used to determine location and magnitude of leaks inside a pressurized water reactor (PWR) containment, as required by plant technical specifications. The major advantage of this detection method is that it minimizes radiation exposure of maintenance personnel because most of the leak detection process is performed from the control room outside the containment. In addition, such a program allows for the elimination of pipe whip restraints and jet impingement shields, eliminating costs for maintenance of these supports and shields in older plants and lowering construction costs for new plants. Previously, only simple single-node containment models were used for determining leakage magnitude. This paper presents a more sophisticated multinode approach for determining the magnitude and location. The resulting sensitivities to leak can be programmed into the ...

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 targeted: the reactor ...

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 reactor proper, the method of ...

International Nuclear Information System (INIS)

Two sodium cooled reactors are currently being operated in the United States of America for the US Department of Energy. These are Experimental Breeder Reactor 11, EBR-11, and the Fast Flux Test Facility, FFTF. EBR-11 is located near Idaho Falls, Idaho, and the FFTF is near Richland, Washington. These reactors are currently engaged in a wide range of testing including fuels and materials tests, and plant system performance and safety development. The US DOE program also includes designs of a next generation sodium cooled power reactor. The FFTF and EBR-11 communities are providing input to these designs. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA.

International Nuclear Information System (INIS)

A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author).

Science.gov (United States)

(accumulates & stores brine for disposal). Separator. (separates water from purge gases). ? Purge pump periodically vent ... Reactor Health. Sensor. ( verifies reactor is operating w/n limits) ... Waste and Hygiene Compartment ...

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 ...

International Nuclear Information System (INIS)

Nuclear physics parameters of the Wolsung CANDU-PHW reactor are computed by use of the PHWCELL computer code that is an improved version of LATREP. The PHWCELL code mainly computes cell parameters of heavy water moderated reactors, and modeling scheme of heavy water reactor cell calculations has been developed with the PHWCELL computer code. The reactor operating conditions considered in the study are cold zero power (CZP) and hot full power (HFP) with equilibrium poison. The cell parameters are also computed as a function of fuel burnup and the numerical results are compared with the results in PSR of the Wolsung unit and in the previous study. (author).

Energy Technology Data Exchange (ETDEWEB)

Water-resources data for the 1993 water year for Minnesota consists of records of stage, discharge and water quality of streams; stage, contents, and water quality of lakes and reservoirs; and water levels and water quality in wells and springs. This volume contains discharge records for 58 gaging stations; stage and contents for 9 lakes and reservoirs; water quality 15 stream stations, 1 lakes station, 1 precipitation station, and water levels for 15 observation wells. Also included are 61 high-flow partial-record stations. Additional water data were collected at various sites, not part of the systematic data collection program and are pulished as miscellaneous measurements.

Energy Technology Data Exchange (ETDEWEB)

In accordance with Public Law 95-217, Section 314 (the 'Clean Lakes' section of the 1977 Amendments to the Federal Water Pollution Control Act), the Massachusetts Division of Water Pollution Control developed a lake classification program based upon the trophic condition of all publicly owned freshwater lakes and ponds in the Commonwealth. This publication, produced and updated on an annual basis since 1976, is the result of that program.

Energy Technology Data Exchange (ETDEWEB)

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outline specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs.

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.

Energy Technology Data Exchange (ETDEWEB)

The verification test program on zinc injection applicability to Japanese BWRs was started in 1997. Laboratory tests using high temperature water loops under BWR reactor water conditions are in progress. This paper is an interim report on results obtained so far. Co-58 and Zn-65 were simultaneously used in the Co radioactivity buildup test to evaluate zinc injection suppression effects towards cobalt deposition on pre-oxidized stainless steel. The following results were obtained. The Co deposition was suppressed effectively by Zn injection, even when there was a pre-oxide film. For the test piping that had the pre-oxide film formed under the NWC (normal water chemistry) condition, when soaked under the HWC (hydrogen water chemistry) condition a large amount of Co-58 was taken into a small part of the inner layer. The distribution ratio of Co-58 in the inner ...

Wastenet

...Corps' tunnel 'expands' understanding of Permafrost Program News Home Warfighter Support Installations Environment Water Resources Information Technology About Us Partnering Opportunities About Us ...News Releases Program News Corps' tunnel 'expands' understanding of Permafrost Corps' tunnel \\

International Nuclear Information System (INIS)

Based on the occurrence of a number of plant incidents during low power and shutdown operating conditions, the Nuclear Regulatory Commission (NRC) has initiated several programs to better quantify risk during these periods. One specific issue of interest is the loss of residual heat removal (RHR) under reduced coolant inventory conditions. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-3 experimental facility at Siemens-KWU to supply applicable data. Recently, an effort has been undertaken at the Idaho National Engineering Laboratory (INEL) to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following loss of vital AC power and consequent loss of the RHR system during reduced inventory operation. The thermal-hydraulic response of a nuclear steam supply system (NSSS) with a closed reactor ...

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 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 operational performance of the FFTF ...

Energy Technology Data Exchange (ETDEWEB)

The purpose of this study is to develop the radioisotope tracer technology, which can be used in solving industrial and environmental problems and to build a strong tracer group to support the local industries. In relation to the tracer technology in 1999, experiments to estimate the efficiencies of a sludge digester of a waste water treatment plant and a submerged biological reactor of a dye industry were conducted. As a result, the tracer technology for optimization of facilities related to wastewater treatment has been developed and is believed to contribute to improve their operation efficiency. The quantification of the experimental result was attempted to improve the confidence of tracer technology by ECRIN program which basically uses the MCNP simulation principle. Using thin layer activation technique, wear of tappet shim was estimated. Thin layer surface of a tappet shim was irradiated by proton beam and the ...

Energy Technology Data Exchange (ETDEWEB)

The five thermal-hydraulic concepts chosen for advanced PWR have been studied as follows: (1) Critical Heat Flux: Review of previous works, analysis of parametric trends, analysis of transient CHF characteristics, extension of the CHF date bank, survey and assessment of correlations, design of a intermediate-pressure CHF test loop have been performed. (2) Passive Cooling Concepts for Concrete Containment system: Review of condensation phenomena with noncondensable gases, selection of a promising concept (i.e., use of external condensers), design of test loop according to scaling laws have been accomplished. and computer programs based on the control-volume approach, and the conceptual design of test loop have been accomplished. (4) Fluidic Diode Concepts: Review of previous applications of the concept, analysis major parameters affecting the performance, development of a computational code, and conceptual investigation of the verification test loop have been ...

Energy Technology Data Exchange (ETDEWEB)

The characterization data obtained to date are described for Approved Testing Material (ATM)-103, which is spent fuel from Assembly D101 of pressurized-water reactor Calvert Cliffs, No. 1. This report is one in a series being written by the Materials Characterization Center (MCC) at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US nuclear waste repository program. ATM-103 consists of 176 full-length irradiated fuel rods with rod-average burnups of about 2600 GJ/kgM (30 MWd/kgM) and less than 1% fission gas release. Characterization data include 1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; 2) isotopic gamma scans; 3) fission gas analyses; 4) ceramography of the fuel and metallography of the cladding; 5) special fuels studies involving analytical transmission ...

Energy Technology Data Exchange (ETDEWEB)

The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the ...

Science.gov (United States)

USGS - science for a changing world USGS Home Contact USGS Search USGS USGS Water-Quality Information * Home * Data * Methods * Labs * Publications * Topics * Programs * Contact Us...

Energy Technology Data Exchange (ETDEWEB)

This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core.

International Nuclear Information System (INIS)

The concept of preliminary transmutation of minor actinides before placement to the long-term storage is considered. The purpose of such preliminary transmutation before ultimate storage is to incinerate a part of actinides and to transform another part into new actinides providing low level of radiotoxicity accumulated in the storage. Modes of transmutation in reactors of PWR, PHWR (CANDU), and Superfenix types are compared. Among power reactors, heavy-water PHWR type reactor is most acceptable for preliminary transmutation. (author)

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)

The ability of the /sup 252/Cf-source-driven neutron noise analysis method to measure subcriticality has been demonstrated in a variety of experimental configurations of fissile materials. Calculations for an approximately 4-m-dia configuration of light water reactor (LWR) fuel elements indicated the feasibility of measuring the subcriticality of large, loosely coupled arrays of LWR fuel elements by this same method. These analysis suggested application to the initial loading of both pressurized and boiling water reactors, zero-power testing of reactors (such as shutdown margin measurements after initial loading), light water reactor refueling, and safe storage of LWR spent fuel. In the fuel storage application, direct measurement of subcriticality in the actual fuel storage facilities provides the parameter which is directly related to ...

Energy Technology Data Exchange (ETDEWEB)

Many mines and processing facilities are in remote, difficult to access areas, or are in areas where access is limited or restricted for national reasons. In a joint project with Canada Centre for Remote Sensing, the Canadian Safeguards Support Program is investigating utility of multi- and hyperspectral remote sensing for remotely inspecting such site. With a view to eventually using the hyperspectral satellite data now coming available, airborne data over uranium and copper mine tailings and uranium processing facilities in northern Canada were acquired. The objective of this work is to demonstrate that multispectral and hyperspectral data can provide complementary and supplementary information to high-resolution panchromatic imagery for the following safeguards applications: Inspection aids (up-to-date maps of remote locations), change detection, evaluation of member state declarations, monitoring of reactor, mining and processing facility ...

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)

In the analysis of the consequences of a hypothetical meltdown accident in an LMFBR, it is important to estimate the final location of the molten fuel pool in the concrete and ground underlying the reactor vessel. The GROWS program and the AYER program have been developed to calculate the final location of the molten fuel pool as the culmination of the transient analysis of this unusual Stefan problem but these programs require extensive computational resources. The solution is provided to the concrete and ground temperatures surrounding the stationary fuel pool and the related heat flux from the pool to the ground surface outside the containment building. This solution can be used to estimate the final location of the fuel pool and to check the end results of the sophisticated programs.

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)

International Nuclear Information System (INIS)

The High Flux Isotope Reactor (HFIR) located at Oak Ridge National Laboratory is one of the world's most powerful research reactors. In 1996, one year after the demise of the Advanced Neutron Source Project, the U.S. Department of Energy embarked on an aggressive program to upgrade the neutron scattering facilities at the HFIR. These upgrades, which are now in progress, include the installation of larger beam tubes, a high-performance hydrogen cold source, and additional neutron guides and neutron scattering instruments. An extensive analysis effort was performed over the past 4 yr to support the design of the modified beamlines and new user facilities and to assess the impact of the upgrades on the integrity of the existing reactor system. The results of three of these analyses are summarized here. Specifically, results are presented for analyses related to the design of the new cold neutron source ...

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 directions and levels of ...

International Nuclear Information System (INIS)

This paper describes the CANDU fuel development activities in Canada during 1997 through 1998. The activities include those of the Fuel Technology Program sponsored by the CANDU Owners Group. The goal of the Fuel Technology Program is to maintain and improve the reliability, economics and safety of CANDU fuel in operating reactors. These activities, therefore, concentrate on the present designs of 28-element and 37-element fuel bundles. The Canadian fuel development activities also include those of the Advanced Fuel and Fuel Cycle Technology Program at AECL. These activities concentrate on the development of advanced fuel designs and advanced fuel cycles, which among other advantages, can reduce the capital and fuelling costs, maintain operating margins in aging reactors, improve natural-uranium utilization, and reduce the amount of spent fuel. (author)

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)

International Nuclear Information System (INIS)

A calculation program (URA 6.F4) was elaborated on FORTRAN IV language, that through finite differences solves the unidimensional scalar Helmholtz equation, assuming only one energy group, in spherical cylindrical or plane geometry. The purpose is the determination of the flow distribution in a reactor of spherical cylindrical or plane geometry and the critical dimensions. Feeding as entrance datas to the program the geometry, diffusion coefficients and macroscopic transversals cross sections of absorption and fission for each region. The differential diffusion equation is converted with its boundary conditions, to one system of homogeneous algebraic linear equations using the box integration technique. The investigation on criticality is converted then in a succession of eigenvalue problems for the critical eigenvalue. In general, only is necessary to solve the first eigenvalue and its corresponding eigenvector, employing ...

International Nuclear Information System (INIS)

This paper outlines the opportunities for UK-Canada collaboration/feasibility studies in areas that include novel research into waste management and decommissioning. A number of Universities in the UK have programs relevant to such collaborations in areas such as fuels; thermal hydraulics, reactor system and materials.

International Nuclear Information System (INIS)

1 - Description of test facility: The LOBI facility is a 1/700 scale model of a four loop PWR and has two primary loops, the intact loop representing three loops and the broken loop representing one loop of a four-loop PWR. The reactor pressure vessel model contains an electrically heated rod-bundle with 64 rods and a heated length of 3.9 m. The nominal heating power is 5.3 MW. The downcomer is of annular shape. An upper head simulator is connected to the vessel. Each of the two primary loops contains a pump and a steam generator. The different mass flows in the loops are established by the pump speeds, since the two pumps are identical. Heat is removed from the steam generators by a secondary system. ECC water can be supplied from two accumulators, one for each loop. Cold or hot leg as well as combined injection can be simulated. The LOBI test facility is the only high pressure integral test facility within the European Communities (1982), ...

Energy Technology Data Exchange (ETDEWEB)

The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

International Nuclear Information System (INIS)

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36 percent to 111 percent of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable.

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)

The paper briefly describes the nuclear reactor facilities at Sandia Laboratories which are used for simulating nuclear weapon produced neutron environments. These reactor facilities are used principally in support of continuing R and D programs for the Department of Energy/Office of Military Application (DOE/OMA) in studying the effects of radiation on nuclear weapon systems and components. As such, the reactors are available to DOE and DOD agencies and their contractors responsible for the radiation hardening of advanced nuclear weapon systems. Emphasis is placed upon two new reactor simulation sources; the Sandia Pulse Reactor-III (SPR-III) Facility which enhances the neutron exposure volume capabilities over those presently available with the existing SPR-II Facility, and the Upgraded Annular Core Pulse Reactor (ACPR) Facility which ...

Science.gov (United States)

From nuclear science symposium; San Francisco, California, USA (14 Nov 1973). A digital Fourier analyzer was programmed to perform reactor neutron noise analysis measurements and on-line processing of the data to obtain the steady-state reactivity. The system is suitable for recovering cross spectral density with low correlatedsignal component and for repetitive measurements with efficient use of reactor time. (auth)

Energy Technology Data Exchange (ETDEWEB)

The mission of the Heavy Water portion of D Area (or 400 Area) at SRS is to purify the site inventory of heavy water for storage in the Reactor Areas for future DOE missions.

International Nuclear Information System (INIS)

Criticality was first achieved with the Fast Flux Test Facility (FFTF) a little more than 10 yr ago on February 9, 1980. Although the FFTF was designed and built primarily for testing fuels, materials, and components for the liquid-metal fast breeder reactor program, it has, over its first 10 yr of operation, provided valuable information in many other areas. This paper provides a summary of the contributions to the physics of liquid-metal reactors (LMRs) obtained from operation of and testing in the FFTF, with emphasis on some of the more significant and interesting accomplishments.

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)

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 input to the individual System Design ...

Energy Technology Data Exchange (ETDEWEB)

The high-purity-oxygen activated sludge process will be used to expand secondary treatment capacity and improve water quality in Santa Monica Bay. The facility is operated by the city of Los Angeles Department of Public Works` Bureau of Sanitation. The overall Hyperion Full Secondary Project is 30% complete, including a new headworks, a new primary clarifier battery, an electrical switch yard, and additional support facilities. The upgrading of secondary facilities is 50% complete, and construction of the digester facilities, the waste-activated sludge thickening facility, and the second phase of the three-phase modification to existing primary clarifier batteries has just begun. The expansion program will provide a maximum monthly design capacity of 19,723 L/s(450 mgd). Hyperion`s expansion program uses industrial treatment techniques rarely attempted in a municipal facility, particularly on such a large scale, including: ...

International Nuclear Information System (INIS)

Appendix B to 10 CFR Part 50 establishes overall quality assurance requirements for the design, construction and operation of safety-related structures, systems, and components. This guide presents a method acceptable to the Commission for complying with these regulations with regard to overall quality assurance program requirements during design and construction of nuclear power plants. Input to this guide has been provided by the Advisory Committee on Reactor Safeguards.

International Nuclear Information System (INIS)

HEDL is conducting an experimental transient testing program to evaluate the performance of prototypic Fast Flux Test Facility (FFTF) fuel pins up to the cladding integrity limit. The relationship is described of the HEDL/TREAT transient overpower test program to the confirmation of FFTF fuel pin design bases via the FFTF fuel pin design procedure.

International Nuclear Information System (INIS)

A brief FFTF project description is provided which includes general plant siting information, general layout, plant design parameters, description of principal systems and components, and description of support facilities. The current status of the FFTF project is provided, including status of plant construction, overall status of the plant checkout and test program, status of operating authorization and plant operating procedures and personnel, and status of reactor core components and experiments. Specific information on the acceptance test program and early program results is discussed. The role of FFTF in the future breeder program is described, including its objectives for verification of plant system and components designs and operability and use as an irradiation test facility.

International Nuclear Information System (INIS)

Purpose: To remove excessive driving pressure applied to an unisolated control rod drive by returning excessive coolant to a condensed water storage tank or to the inlet side of a drive water pump using a coolant flow rate control pipe of a control rod driving hydraulic system. Constitution: Excessive water is returned to a condensed water tank while controlling the excessive coolant by a flow control valve in response to variations in the pressure difference between the reactor pressure and the driving water line when the control rods are isolated using a pipe from the outlet side of the drive water pump to the condensed water storage tank. Thus, the control rod to be isolated is prevented form being dropped. (Sekiya, K.).

Energy Technology Data Exchange (ETDEWEB)

Nuclear power has been considered for space applications since the 1960s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors/ rocket-engines in the Rover/NERVA programs. However, changes in environmental laws may make the redevelopment of the nuclear rocket more difficult. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel form significantly different from NERVA may be needed to ensure public support. The Center for Space Nuclear Research (CSNR) is pursuing development of tungsten based fuels for use in a NTR, for a surface power reactor, and to encapsulate radioisotope power sources. The CSNR Summer Fellows program has investigated the feasibility of several missions enabled by the NTR. The potential mission benefits of a nuclear rocket, historical achievements of the previous programs, and recent ...

Energy Technology Data Exchange (ETDEWEB)

In light of the recent rapid increase in the fossil fuel prices it is meaningful to evaluate the impact of these price changes in the economics of dual-purpose desalination projects producing electricity and fresh water simultaneously. The price of crude oil and LNG (Liquefied Natural Gas) has increased by about 200% and 100% during the past three or four years. The uranium price has also increased by nearly 500% during the same period. The purpose of this paper is to analyze and compare the economics of SMART (System-integrated Modular Advanced ReacTor) which is being developed as a small size PWR type and the LNG Combine Cycle coupled with MED (Multi-Effect Distillation) which are being acknowledged as promising energy sources for the future in Korea. The methods of analysis used in this paper are the lifetime leveled cost method for the power and water cost calculation and the power credit method for the total cost ...

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 ...

International Nuclear Information System (INIS)

This paper describes the Fast Flux Test Facility (FFTF) Operations initial training program utilizing the Operator Training Simulator (OTS). The OTS is a computer-driven system that provides real time response of essential FFTF plant functions to a control room mockup. The FFTF, a 400 Megawatt, three-loop, sodium-cooled fast test reactor will test fuels, materials and equipment for the U.S. Liquid Metal Fast Breeder Reactor Program. Construction is expected to be completed in August 1978. Initial criticality is expected in early 1979. This schedule will require FFTF control room operators to be fully qualified to operate the facility by late 1979. Because FFTF is like no other U.S. nuclear reactor, existing U.S. utility plants could not be depended on to provide highly experienced people to operate FFTF. Therefore, an Operator Training Simulator has been built. The OTS will play a ...

Science.gov (United States)

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.

Energy Technology Data Exchange (ETDEWEB)

Water injection into a vapor-dominated geothermal reservoir is an effective method of sustaining steam production from the field. Injection puts additional water to the reservoir and raises the prevailing reservoir pressure. This process improves the field`s productivity. However, the increased pressure also increases the water retention capacity of the reservoir rocks through the effects of adsorption and capillary condensation. Due to the significant costs associated with water injection programs, optimizing injection not only involves maximizing the energy yield from the resource but also the present worth of the project. Two crucial parameters that need to be established are: (1) how much to inject; and, (2) when to inject it. This study investigated the optimal design of these parameters. It was found that comparable energy yield can be attained for injection ...

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)

Record production of heavy water was achieved: the plants at Glace Bay and Port Hawkesbury, Nova Scotia, produced a total of 560 megagrams. A shipment of 500 Mg was delivered on time to the Wolsung CANDU reactor in Korea. Energy conservation and waste hea...

Energy Technology Data Exchange (ETDEWEB)

The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making ...

Science.gov (United States)

In October 1980, Air Products and Chemicals, Inc. began a three year contract with the DOE: Catalyst and Reactor Development for a Liquid Phase Fischer-Tropsch Process. The program contains four major tasks: (1) Project Work Plan, (2) Slurry Catalyst Development, (3) Slurry Reactor Design Studies, and (4) Pilot Facility Design. This report describes work on Tasks 2 and 3 carried out in the third quarter of the contract. In Task 2, the computerized search of the Fischer-Tropsch literature was continued, and improvements were made in data processing programs. Shakedown tests were completed on the first 300 ml slurry reactor, and construction of the second and third reactors began. Five modified conventional slurry catalysts were prepared, and two batches were tested in the gas phase giving information on selectivity as a function of composition and activation. ...

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.

International Nuclear Information System (INIS)

Existing applications of Alloy 800 are summarized, with particular reference to its use in various types of reactor. The need for a co-ordinated research and development programme is stressed, and the variables to be explored are outlined. The papers relating to the problem of corrosion and cracking in water and steam are considered. the strength and ductility of Alloy 800 is considered. Finally, sections of the summary deal with the use of Alloy 800 for (a) sodium cooled fast reactor boiler tubes; (b) the high temperature gas cooled reactor; and (c) PWR steam generator tubes. (U.K.).

International Nuclear Information System (INIS)

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)

Problem areas in BWRs, PWRs and PHWRs, from the viewpoint of chemistry, and the problem of fission product release in nuclear reactors are discussed. These problem areas are : fuel performance, off-normal water chemistry due to condenser leaks, the transport and deposition of the activated corrosion and fission products, denting in steam generators (in the case of PWRs), ingress of air in the cover gas helium and consequent radiolysis of D_2O in the moderator circuit (in the case of PHWRs). (M.G.B.).

International Nuclear Information System (INIS)

On the 3. and 4. November 1982 the sixth conference of the Corporation for Reactor Safety (GRS) was held in Cologne's Guerzenich. The theme of this year's meeting was the 'Status of Risk Investigations at Nuclear Power Plants'. A principal topic was a report on findings made by the GRS during the 'Risk Oriented Analysis SNR-300'. The second topic comprised the newest developments within Phase B of the Risk Study of Water Pressure Reactors, the discussion of the dose/effect relationship and considerations on threshold risk values. (orig.).

International Nuclear Information System (INIS)

SMART (System integrated Modular Advanced ReacTor) , is under development at the Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection system (SIS), and an adoption of 4 trains of passive residual heat removal system (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a steam generator tube rupture (SGTR) is one of the most important initiating events which results in a high core damage frequency. Clear understanding of accident ...

Energy Technology Data Exchange (ETDEWEB)

Description of the current status of the developments of a simulation of the Darlington Nuclear Generating Station shutdown and regulating systems, DARSIM. The DARSIM program simulates the spatial neutron dynamics, the regulations of the reactor power, and shutdown system 1 and shutdown system 2 software. The DARSIM program operates in the interactive simulation (INSIM) program environment. DARSIM was installed on the APOLLO computer at the Atomic Energy Control Board (AECB) and a version for an IBM-PC was also provided for the exclusive use of the AECB. Shutdown system software was updated to incorporate the latest revisions in the functional specifications. Additional developments were provided to assist in the use and interpretation of the DARSIM results.

International Nuclear Information System (INIS)

The basic Quality Assurance Program Requirements standard for the National Waste Terminal Storage Program has been developed primarily for nuclear reactors and other fairly well established nuclear facilities. In the case of waste isolation, however, there are many ongoing investigations for which quality assurance practices and requirements have not been well defined. This paper points out these problems which require supplementary requirements. Briefly these are: (1) the language barrier, that is geologists and scientists are not familiar with quality assurance (QA) terminology; (2) earth sciences deal with materials that cannot be characterized as easily as metals or other materials that are reasonably homogeneous; (3) development and control of mathematical models and associated computer programs; (4) research and development.

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).

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 density, which will also allow the conversion ...

International Nuclear Information System (INIS)

This study evaluates the degree to which surface impoundments at licensed facilities comply with significant changes in NRC requirements initiated by enactment of EPA's final environmental standards for uranium recovery facilities (40 CFR Part 192). Impoundment liner requirements, ground-water protection standards, ground-water monitoring and corrective action programs, and site closure standards are the most significant regulatory modifications. The compliance status of 30 conventional mills and 31 in-situ mines is determined from a review of Nuclear Regulatory Commission and agreement state docket files through November, 1983. Results of this review show that a majority of conventional uranium mill tailings management systems are deficient with respect to liner requirements for new impoundments or proposed expansions to existing impoundments, as well as with respect to some aspects of ground-water ...

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 decreasing ...

International Nuclear Information System (INIS)

MOX fuel has been one of the mature nuclear fuels which can be used in light water nuclear power plant now. The development status in this domain in foreign countries, the major influence of MOX fuel on reactor performance and the countermeasures are introduced in this paper. The application of MOX fuel in China's PWR is discussed in the end. (authors)

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.

International Nuclear Information System (INIS)

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power ...

International Nuclear Information System (INIS)

The Swedish State Power Board has together with Nukem, Hanau, West-Germany carried out pyrolysis o powder resins in a pilot plant with a capacity of about 30 kg/hr. The pyrolysis reactor with its afterburner and offgas scrubber system has been operated under steady state condition. About 2200 kg resins have been pyrolysed under November-December 1983 and the decontamination factor for Cs has been measured. Solidification of the residues from the pyrolysis reactor and scrubber water solutions has been carried out and various recipes with cement have been tested. The pyrolysis process has high decontamination factors and no offgas problems as the operating temperature of the reactor is Low. The residues from the reactor are chemically dead and can not cause swelling problems. Compared with a normal cementation process the final waste volume will be reduced with a factor of 4 if also ...

Energy Technology Data Exchange (ETDEWEB)

The 3001 Storage Canal is located under portions of Buildings 3001 and 3019 at Oak Ridge National Laboratory (ORNL) and has a capacity of approximately 62,000 gallons of water. The term canal has historically been used to identify this structure, however, the canal is an in-ground reinforced concrete structure satisfying the regulatory definition of a tank. From 1943 through 1963, the canal in Building 3001 was designed to be an integral part of the system for handling irradiated fuel from the Oak Ridge Graphite Reactor. Because one of the main initial purposes of the reactor was to produce plutonium for the chemical processing pilot plant in Building 3019, the canal was designed to be the connecting link between the reactor and the pilot plant. During the war years, natural uranium slugs were irradiated in the reactor and then pushed out of the graphite matrix into the system of ...

British Library Electronic Table of Contents (United Kingdom)

A dense Pd-Ag membrane reactor (MR) with 100% hydrogen selectivity packed with either Rh/La2O3 or Rh/La2O3-SiO2 as catalysts was used to carry out the dry reforming of methane. The membrane reactor simulation was performed using a well-known reactor model. For this purpose, we employed the equations derived from complete kinetic studies of the dry reforming of methane reaction in connection with both catalysts. In addition, we developed the kinetic equation for the reverse water gas shift reaction (RWGS). The combination of detailed kinetic studies with the measured permeation flux for the Pd-Ag membrane allowed a complete comparison between experimental and simulated operation variables. The variables studied for both catalysts were methane conversion and hydrogen permeation as a function...

Energy Technology Data Exchange (ETDEWEB)

Activities are reported by the Reactor Systems Section, Research Instrument Section, and the Measurement and Controls Engineering Section. Reactor system activities include dynamic analysis, survillanc and diagnostic methods, design and evaluation, detectors, facilities support, process instrumentation development, and special assignments. Activities in the Research Instrument Section include the Navy-ORNL RADIAC development program, advanced ..gamma.. and x ray detector systems, neutron detection and subcriticality measurements, circuit development, position-sensitive detectors, stand-alone computers, environmental monitoring-detectors and systems, plant security, engineering support for fusion energy division, engineering support for accelerator physics, and communications: radio, closed-circuit tv, and computer. Activities in the Measurement and Controls Engineering Section include the AVLIS program; ...

International Nuclear Information System (INIS)

This paper presents an overview of issues related to surface water contamination arising from uranium mining activities in the Alligator Rivers Region (ARR) of northern Australia, and a program of research and monitoring that must assess the potential and actual effects on ecosystem and human health. The program of assessing effects on aquatic ecosystems involves a four-tiered approach including the derivation of local water quality guideline trigger values, direct toxicity assessment of mine waters prior to their release, creekside or in situ monitoring for early warning of adverse effects during mine water release, and longer-term monitoring of macroinvertebrate and fish communities. Bioaccumulation in aquatic biota is also assessed, and is an issue of importance not only to ecosystem health, but also to the health of local Aboriginal people. The aquatic ...

Energy Technology Data Exchange (ETDEWEB)

An important component of the overall program to validate WIMS-AECL for use with RFSP in the analysis of CANDU-6 reactors for design and safety analysis calculations is the validation of calculations of incremental cross sections used to represent reactivity devices. A method has been developed for the calculation of the three-dimensional neutron flux distribution in and around CANDU reactor fuel channels and reactivity control devices. The methods is based on one- and two dimensional transport calculations with the WIMS-AECL lattice cell code, SPH homogenization, and three-dimensional flux calculations with finite-difference diffusion theory using the MULTICELL code. Simulations of Wolsung 1 Phase-B commissioning measurements and Point Lepreau restart tests have been performed, as a part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor ...

British Library Electronic Table of Contents (United Kingdom)

Spherical nano-sized YSZ (yttria stabilized ZrO2) powders were successfully synthesized via a reverse microemulsion system. The water droplets in the microemulsion system of yclohexane/water/span85/Triton X-100/hexyl alcohol can act as the nano-reactors which solubilize zirconium oxychloride and ammonia water separately. The minute original reactors are favor to the formation of nano-sized spherical YSZ powders and the dispersibility of the powders can be controlled effectually by adjusting the weight ratio of the LiNO3 molten salt to the precursor. The phase transformation from cubic to monoclinic starts at and 500??C and finally monoclinic and cubic phase with increased crytallinity coexist at 800??C. The effect of LiNO3 molten salt in the formation of YSZ powders was also discussed.

International Nuclear Information System (INIS)

Purpose: To provide a feedwater control device capable of minimizing the adverse response of steam drum level at low power. Consitution: In order to perform feedwater control at low power by the substantial control of three factors, that is, main steam flow rate, feedwater flow rate and steam drum level, the main steam flow rate is determined from the reactor output and feedwater rate is determined from the changes in the feedwater temperature due to the mixing of waters in the reactor clean up system and feedwater. If a difference is resulted between these flow rates, a starting feedwater regulator is controlled instantly to eliminate the difference. The water level in the steam drum is used for amending the difference from the final set value of the drum water level, by which the adverse response of the steam drum level can be minimized. (Seki, T.).

International Nuclear Information System (INIS)

The aim of this work is the implantation and characterization of a neutron radiography system that uses an electronic device for attainment of images in real time, for its implementation in the nuclear research reactor Argonauta at IEN/CNEN (Nuclear Engineering Institute of the Brazilian Nuclear Energy Commission). The Electronic Imaging System in Real Time is composed by a scintillator screen for neutron, a video camera (CCD), a digital plate and a computer with specific computational programs for digital processing of the images. The System in installed real time is apt to carry through neutron radiography inspections of static and dynamic events of several types of samples. (author)

Energy Technology Data Exchange (ETDEWEB)

The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through September 1999.

International Nuclear Information System (INIS)

Results of calculational investigations into circulation of water and steam-and-water mixture in the PGV-1000 steam generator heat exchanger bundle used at NPPs with the WWER-1000 reactors, are considered. Model of water circulation in horizontal steam generator with submerged heating surface under conditions of steam generation irregularity along the heat transfer tubes is made. On the basis of the obtained data the assumption is made about water essential overflows from the hot collector zone into the cold one. Overflow rate over the upper line of the heat transfer tubes may constitute 0.7 m/s. The conclusion is made about the necessity to set up the vertical barrier which divides hot and cold sections of heat transfer tubes and helps to avoid water transverse overflows.

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 ...

Energy Technology Data Exchange (ETDEWEB)

Waste management, like many other issues, has experienced major milestones. In 1971, the Calvert Cliff's decision resulted in an entirely different approach to the consideration of environmental impact analysis in reactor siting. The accidents at Three Mile Island and Chernobyl have had profound effects on nuclear power plant design. The high-level waste repository program has had many similar experiences that have modified the course of events. The management of radioactive, hazardous chemical and mixed waste in all of the facilities of the Oak Ridge Operations (ORO) Office of the Department of Energy (DOE) took on an entirely different meaning in 1984. On April 13, 1984, Federal Judge Robert Taylor said that DOE should proceed 'with all deliberate speed' to bring the Y-12 plant into compliance with the Resource Conservation and Recovery Act and the Clean Water Act. This decision resulted from a ...

International Nuclear Information System (INIS)

Tarapur Atomic Power Station Unit-3 and 4 (TAPS -3 and 4) are the 540 MWe reactors. Unit-4 attained first criticality on 06th March 2005 and operated for about 230 effective full power days (EFPD). Unit-3 attained first criticality on 21st May 2006 and operated for about 20 EFPD. With the reactor operation radiation field increases on the Primary Heat Transport system equipments, Moderator system equipments and auxiliary system equipments due to deposition of fission products and activation products in different reactor systems. These dose rates significantly contributes to the external exposure and stations collective dose during reactor operation, refueling operation and maintenance activities. A study was undertaken at TAPS 3 and 4 to identify the system equipments showing the significant dose rates and identify the radionuclides present in the primary heat transport system, Moderator systems, cover ...

Energy Technology Data Exchange (ETDEWEB)

This report is a summary of the progress and research produced for the Idaho National Engineering and Environmental Laboratory (INEEL) Boron Neutron Capture Therapy (BNCT) Research Program for calendar year 1996. Contributions from the individual investigators about their projects are included, specifically, physics: treatment planning software, real-time neutron beam measurement dosimetry, measurement of the Finnish research reactor epithermal neutron spectrum, BNCT accelerator technology; and chemistry: analysis of biological samples and preparation of {sup 10}B enriched decaborane.

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.

Energy Technology Data Exchange (ETDEWEB)

This report discusses environmental surveillance activities conducted on the Nevada Test Site during the first calendar quarter of 1996. Surveillance activities included collection and analysis of air, noble gas, tritiated water vapor, and other water samples. Samples were analyzed for both radiological and non-radiological effluents.

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 east), Deep River (10 km west) and Chalk ...

International Nuclear Information System (INIS)

Elevated pH operation to the pH value of 7.3 at 285degC is known to be effective for the reduction of radiation source in the primary water system of PWRs. A research project was started in 1989 and concluded in 1996 to study and verify the optimum pH and/or Li concentration from the viewpoint of radiation source reduction and materials integrity under improved water chemistry. This research project is sponsored by the Ministry of International Trade and Industries (MITI) in Japan and has two programs; high pH and high Li. The high Li program was conducted to establish the optimum Li concentration for the high boron concentration region (1100 - 1800 ppm) of the high burn up operation. In this paper, we shall discuss radiation source behavior under high pH conditions and PWSCC (Primary Water Stress Corrosion Cracking) susceptibility of materials with change of primary ...

Energy Technology Data Exchange (ETDEWEB)

Because of the importance of water vapor, the ARM program initiated a series of three intensive operating periods (IOPs) at its CART (Cloud And Radiation Testbed) site. The goal of these IOPs is to improve and validate the state-of-the-art capabilities in measuring water vapor. To date, two of the planned three IOPs have occurred: the first was in September of 1996, with an emphasis on the lowest kilometer, while the second was conducted from September--October 1997 with a focus on both the upper troposphere and lowest kilometer. These IOPs provided an excellent opportunity to compare measurements from other systems with those made by the CART Raman lidar. This paper addresses primarily the daytime water vapor measurements made by the lidar system during the second of these IOPs.

International Nuclear Information System (INIS)

1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type ...

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 ...

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 ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

It is needless to say that one of the most important roles of the condensate demineralizer in Japanese boiling water reactors (BWR) is to eliminate such impurities during accidental occurrence of sea water leakage from condensate cooling system. Ion exchange resins packed in condensate demineralizer have also been expected to decrease crud, or corrosion products (CP) in condensate water in order to finally reduce activated corrosion products (ACP) in the reactor coolant loop. It is perceived that crud removal ability of a condensate demineralizer has been improved year by year. And we call this phenomenon as `Aging Effect`. Typical property changes of aged cation exchange resin consisted of an increase of water retention capacity and a change of surface texture. Based on these findings, we formulated a new concept and developed new gel type ion exchange resins ...

International Nuclear Information System (INIS)

In General Electric's design of the Simplified Boiling Water Reactor (SBWR), The depressurization valves (DPV) are installed in the reactor pressure boundary: four are connected to the reactor vessel by means of nozzles, and two more are located on the main steam pipes (one DPV for each line), which act during particular transients and/or loss of coolant accidents (LOCA), consequently providing the reactor vessel with a safe quick depressurization system. Once the vessel is de pressurised, the passive gravity-driven cooling system (GDCS) starts to operate, permitting the injection of water required for continuous core cooling. DPVs are leak tight, with welded flaps, actuated by a [striker[hammer***] which is activated by an explosive mixture. The dynamic loads that open these valves include, in addition to those produced by steam (typical in any thermodynamic ...

Energy Technology Data Exchange (ETDEWEB)

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for ...

International Nuclear Information System (INIS)

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for ...

International Nuclear Information System (INIS)

Purpose: To effectively cool the reactor core in a steam atmosphere by upwardly directing several of spray nozzles attached to a ring header thereby increasing the flying distance of the spray. Constitution: Ring headers in two upper and lower stages are disposed above the outer circumference of a reactor core and each of the ring headers is mounted with spray nozzles. Among the spray nozzles, at least several nozzles mounted to the ring header at the lower stage are directed such that the center axis for each of the nozzle is raised above the horizontal axis and other several nozzles are mounted with the nozzle center axis directed downwardly from the horizontal axis. Accordingly, even if collapsing phenomenon occurs in the jetting stream due to the condensation in the steams that forms the operation atmosphere of the reactor core spray cooling device, a sufficient amount of emergency cooling water can ...

International Nuclear Information System (INIS)

Scenarios are developed for spent fuel that is taken out of the nuclear reactor and stored for 20 years before reprocessing, and three storage techniques which use a water pool, dry cask or vault are compared with respect to their costs. The storage price (storage cost per kilogram of spent fuel), which is employed as the economic index, is calculated on the assumption that all the charge is paid when the spent fuel is brought in the storage facilities. Four scenarios are assumed for spent PWR and BWR fuels to be stored in at-reactor (AT) or away-from-reactor (AFR) facilities. The capital costs cover the buildings of the storage facilities, equipment, decommissioning, casks for transportation and storage (for cask storage) and casks for transportation (for water pool or vault storage). Costs for operation and maintenance of the facilities are also considered in evaluating these ...

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 input to the individual System Design ...

International Nuclear Information System (INIS)

There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase ...

International Nuclear Information System (INIS)

Nuclear facilities of the Department of Energy (DOE) located at the Savannah River Site must comply with DOE orders as implemented at DOE-SR. The DOE orders cover safety criteria, design criteria, environmental protection, occupational health and safety. The program applies to DOE and contractors. In this section, the Nuclear Regulatory Commission (NRC) criteria and industry codes and standards are addressed as well as DOE orders. Specific DOE orders which add additional criteria have also been noted. A program for assessing and implementing contractor applicable DOE orders has been established. This program ensures that compliance is achieved through developing and implementing policies, programs, and procedures. The primary emphasis is placed on safe, efficient reactor restart and operation. DOE has classified orders applicable to restart as Level I, Category I while those ...

International Nuclear Information System (INIS)

As the Operating and Support staffs of the FFTF organization have gained experience, the plant reliability and capacity factors have shown a steadily improving trend. The plant capacity factor for Cycle 4 was 99.5%. It is the purpose of this report to describe the evolution of the maintenance organization at the FFTF site from a general support organization to a technically proficient organization playing a major role in planning and performance of plant maintenance evolutions.

Science.gov (United States)

A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters.

Energy Technology Data Exchange (ETDEWEB)

Shield measurements and four cycles of operating experience have shown the design and construction of radiation shields for the Fast Flux Test Facility (FFTF) reactor and plant to be satisfactory. A number of minor shield deficiencies were found and corrected. Most of these were associated with interfaces between components, each of which was satisfactory by itself. Preliminary evaluation of the shield measurements indicates satisfactory agreement with design calculations. Operator doses to date have been quite small, especially when compared to light water reactor experience.

Energy Technology Data Exchange (ETDEWEB)

Recently the real-time neutron radiography system of the Kyoto University Reactor (KUR) has been developed and practically applied to penetrating the side plates of the MTR type reactor fuels and investigation of moving objects. In this paper an application of the KUR neutron TV system to neutron computed tomography (NCT) is described. By using the NTV system, projection data can be acquired in a single measurement and simultaneously the projection image can be observed on a CRT monitor. The Fourier-convolution technique is used to produce the reconstructed image and its image has a good enough quality for revealing water in a small hole of 1.5 mm in diameter.