Plant maintenance and advanced reactors, 2006

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal (ed.)

2006-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

Marine reactor pressure vessels dumped in the Kara Sea

International Nuclear Information System (INIS)

Mount, M.E.

1997-01-01

Between 1965 and 1988, 16 marine reactors from seven Russian submarines and the icebreaker Lenin, each of which suffered some form of reactor accident, were dumped in a variety of containments, using a number of sealing methods, at five sites in the Kara Sea. All reactors were dumped at sites that varied in depth from 12 to 300 m and six contained their spent nuclear fuel (SNF). This paper examines the breakdown of the reactor pressure vessel (RPV) barriers due to corrosion, with specific emphasis on those RPVs containing SNF. Included are discussions of the structural aspects of the steam generating installations and their associated RPVs, a summary of the disposal operations, assumptions on corrosion rates of structural and filler materials, and an estimate of the structural integrity of the RPVs at the present time (1996) and in the year 2015

Intelligent information database of the thermal-hydraulic characteristics for a future marine water reactor

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki

2000-01-01

At the Ship Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated type marine water reactor has been conducted. This current study aims at developing an intelligent information database program with the thermal-hydraulic characteristics of a future marine water reactor on the basis of the valuably experimental knowledge, which was obtained from the above-mentioned studies. In this paper, the experimental knowledge with the flow boiling of a once-through steam generator and the natural circulation of primary water under a ship rolling motion was converted into an intelligent information database program. The program was created as a Windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability for any helical-coil type once-through steam generator design, (2) reference and graphic display of the experimental data, (3) reference of the information such as analysis method and experimental apparatus. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized reactor with helical-coil type steam generator. (author)

Design and properties of marine reactors and associated R and D

Energy Technology Data Exchange (ETDEWEB)

Gagarinski, A; Ignatiev, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs.

Design and properties of marine reactors and associated R and D

International Nuclear Information System (INIS)

Gagarinski, A.; Ignatiev, V.; Devell, L.

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs

Research on improvement of marine nuclear reactors and future perspective

International Nuclear Information System (INIS)

Yokomura, Takeyoshi

1988-01-01

The features when atomic energy is utilized on the sea are that the fuel cost is low, accordingly it is suitable to the power sources of large output, that the volume and weight of fuel are small, accordingly it is suitable to the continuous operation for a long period without refueling, and that oxygen is not required for the burning, accordingly it is suitable to undersea power sources. In USSR, four nuclear icebreakers have been in use, and four more are under construction. A nuclear LASH ship has been operated, and one more is under construction. As the other fields than sea transportation, an electricity generation barge MH-1A of USA used as the auxiliary power source for the Panama Canal and a research submarine NR-1 of USA have been in practical use. With the advance of ocean development in future, the creation of needs such as deep sea power stations, deep sea research ships and deep sea work ships is expected. Marine nuclear reactor technology was begun in the form of the nuclearization of merchant ships, and Savannah of USA, Otto Hahn of West Germany and Mutsu of Japan were built. The marine nuclear reactors built so far and of which the conceptual design was carried out are shown. The improvement of marine reactors is the reduction of size and weight, the simplification of the system, the adoption of self pressurization and self compensation and so on. The research on the improvement in Japan Atomic Energy Research Institute is reported. (Kako, I.)

Simplified simulation of an experimental fast reactor plant

International Nuclear Information System (INIS)

Fujii, Masaaki; Fujita, Minoru.

1978-01-01

Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

Evaluation of the trial design studies for an advanced marine reactor, (2)

International Nuclear Information System (INIS)

Ambo, Noriaki; Yokomura, Takeyoshi.

1988-03-01

As for the CARAMEL fuel (plate-type fuel) that was the fuel of the integrated-type reactor which was one of the trial design studies for an Advanced Marine Reactor, its structure and its fuel specific characteristics were studied and compared with a fuel rod (cylindrical fuel), and the total characteristics of the caramel fuel was reviewed and evaluated. (author)

U.S. Nuclear Power Reactor Plant Status

Data.gov (United States)

Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

Fast reactor system factors affecting reprocessing plant design

International Nuclear Information System (INIS)

Allardice, R.H.; Pugh, O.

1982-01-01

The introduction of a commercial fast reactor electricity generating system is very dependent on the availability of an efficient nuclear fuel cycle. Selection of fuel element constructional materials, the fuel element design approach and the reactor operation have a significant influence on the technical feasibility and efficiency of the reprocessing and waste management plants. Therefore the fast reactor processing plant requires liaison between many design teams -reactor, fuel design, reprocessing and waste management -often with different disciplines and conflicting objectives if taken in isolation and an optimised approach to determining several key parameters. A number of these parameters are identified and the design approach discussed in the context of the reprocessing plant. Radiological safety and its impact on design is also briefly discussed. (author)

Nuclear power plant with several reactors

Energy Technology Data Exchange (ETDEWEB)

Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

1972-05-10

A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

Improving plant availability by predicting reactor trips

International Nuclear Information System (INIS)

Frank, M.V.; Epstein, S.A.

1986-01-01

Management Ahnalysis Company (MAC) has developed and applied two complementary software packages called RiTSE and RAMSES. Together they provide an mini-computer workstation for maintenance and operations personnel to dramatically reduce inadvertent reactor trips. They are intended to be used by those responsible at the plant for authorizing work during operation (such as a clearance coordinator or shift foreman in U.S. plants). They discover and represent all components, processes, and their interactions that could case a trip. They predict if future activities at the plant would cause a reactor trip, provide a reactor trip warning system and aid in post-trip cause analysis. RAMSES is a general reliability engineering software package that uses concepts of artificial intelligence to provide unique capabilities on personal and mini-computers

Discharges from a fast reactor reprocessing plant

International Nuclear Information System (INIS)

Barnes, D.S.

1987-01-01

The purpose of this paper is to assess the environmental impact of the calculated routine discharges from a fast reactor fuel reprocessing plant. These assessments have been carried out during the early stages of an evolving in-depth study which culminated in the design for a European demonstration reprocessing plant (EDRP). This plant would be capable of reprocessing irradiated fuel from a series of European fast reactors. Cost-benefit analysis has then been used to assess whether further reductions in the currently predicted routine discharges would be economically justified

The water desalination complex based on ABV-type reactor plant

International Nuclear Information System (INIS)

Panov, Yu.K.; Fadeev, Yu.P.; Vorobiev, V.M.; Baranaev, Yu.D.

1997-01-01

A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

Nuclear reactor kinetics and plant control

CERN Document Server

Oka, Yoshiaki

2013-01-01

Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

Evaluation of the trial design studies for an advanced marine reactor, (3)

International Nuclear Information System (INIS)

Ambo, Noriaki; Yokomura, Takeyoshi.

1988-03-01

JAERI have carried out four core designs for three different type reactors (Semi-Integrated, Integrated and Integrated (self-pressured) type reactors), as the trial designs of an Advanced Marine Reactor for three years (1983 ∼ 1985). This report describes the result of comparison and studies of the core specific characteristics of these four cores, which include core concept, specifications, core life, specific power density, burn-up, reactivity control and etc. In conclusion, it was found that the Integrated type reactor core and the Semi-Integrated type reactor core designs satisfy the conditions of long core life (four years), high specific power density (50 ∼ 61 kw/l) and high burn-up (30,000 ∼ 32,000 MWD/t), so these two cores will be optimum designs based on the present technologies. (author)

Engineering design of advanced marine reactor MRX

International Nuclear Information System (INIS)

1997-10-01

JAERI has studied the design of an advanced marine reactor (named as MRX), which meets requirements of the enhancement of economy and reliability, by reflecting results and knowledge obtained from the development of N.S. Mutsu. The MRX with a power of 100 MWt is intended to be used for ship propulsion such as an ice-breaker, container cargo ship and so on. After completion of the conceptual design, the engineering design was performed in four year plan from FY 1993 to 1996. (1) Compactness, light-weightiness and simplicity of the reactor system are realized by adopting an integral-type PWR, i.e. by installing the steam generator, the pressurizer, and the control rod drive mechanism (CRDM) inside the pressure vessel. Because of elimination of the primary coolant circulation pipes in the MRX, possibility of large-scale pipe break accidents can be eliminated. This contributes to improve the safety of the reactor system and to simplify the engineered safety systems. (2) The in-vessel type CRDM contributes not only to eliminate possibilities of rod ejection accidents, but also to make the reactor system compact. (3) The concept of water-filled containment where the reactor pressure vessel is immersed in the water is adopted. It can be of use for emergency core cooling system which maintains core flooding passively in case of a loss-of-coolant accident. The water-filled containment system also contributes essentially light-weightness of the reactor system since the water inside containment acts as a radiation shield and in consequence the secondary radiation shield can be eliminated. (4) Adoption of passive decay heat removal systems has contributed in a greater deal to simplification of the engineered safety systems and to enhancement of reliability of the systems. (5) Operability has been improved by simplification of the whole reactor system, by adoption of the passive safety systems, advanced automatic operation systems, and so on. (J.P.N.)

Enrichment of anammox bacteria fro marine environment for the construction of a bioremediation reactor

Energy Technology Data Exchange (ETDEWEB)

Nakajima, Jun; Sakka, Makiko; Kimura, Tetsuya; Sakka, Kazuo [Mie Univ., Tsu (Japan). Graduate School of Bioresources; Furukawa, Kenji [Kumamoto Univ. (Japan). Dept. of Civil Engineering and Architecture

2008-01-15

In the global ocean nitrogen cycle, the anaerobic ammonium-oxidizing (anammox) process is recognized as important. In this study, we established an enrichment culture of marine anammox bacteria (MAB) in a column-type reactor. The reactor, which included a porous polyester non-woven fabric that had been placed at the sea floor in advance for enrichment, was continuously fed with NH{sub 4}Cl and NaNO{sub 2} for more than 1 year. Anammox activity in the MAB reactor was confirmed by {sup 15}N tracer analysis using {sup 15}NH{sub 4}Cl and Na{sup 14}NO{sub 2}. We identified two 16S rRNA genes in the amplified DNA fragments derived from MAB, which were highly homologous with those from Candidatus ''Scalindua wagneri'' and an uncultured planctomycete clone. Fluorescence in situ hybridization analysis using an anammox-specific probe also confirmed that MAB predominated in the reactor. To our knowledge, this is the first report on the establishment of an enrichment culture of anammox bacteria from the marine environment using a continuous culture system. (orig.)

Adapting a reactor safety assessment system for specific plants

International Nuclear Information System (INIS)

Ballard, T.L.; Cordes, G.A.

1991-01-01

The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

Seaweed: A valuable marine plant source

Digital Repository Service at National Institute of Oceanography (India)

Dhargalkar, V.K.; Pereira, N.

Seaweed, one of the components of seashore, is the most unique and important marine living resource. Its multi-ranged applications are of economic as well as ecological importance. These plants have been a source of food, feed and medicine...

Advanced designs of VVER reactor plant

International Nuclear Information System (INIS)

Mokhov, V.A.

2010-01-01

The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

Gas core reactor power plants designed for low proliferation potential

International Nuclear Information System (INIS)

Lowry, L.L.

1977-09-01

The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

Marine biomass power plant using methane fermentation

Energy Technology Data Exchange (ETDEWEB)

Matsui, T.; Saito, H.; Amano, T.; Sugawara, H.; Seki, T.; Abe, T. [Technology Research Inst., Tokyo Gas Co. Ltd., Tokyo (Japan)

2004-07-01

This study presented an effective way to produce biogas from the large quantities of seaweed waste in Japan. A large-scale marine biomass pilot plant was built to produce biogas from marine biomass. Methane fermentation was the process used to produce biogas from Laminaria sp. The maximum treating capacity of the pilot plant is 1 ton of seaweed per day. The pilot plant includes a pretreatment facility, fermentation, biogas storage and power generation. The maximum methane yield from the biomass plant is 22 cubic ton-seaweed. The purified biogas has generated 10 kW of electricity and 23 kW of heat. The biogas was also mixed with natural gas for use in a gas engine generator. The engine operation remained stable despite changes in quantity and composition of the collected biogas caused by changes with the source of biomass and sea conditions. The thermal efficiency of the gas engine running on mixed biogas and natural gas was more than 10 per cent higher than an engine running on biogas fuel alone. 4 refs., 2 tabs., 3 figs.

PWR type reactor plant

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1993-01-01

A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

Analysis of reactor trips involving balance-of-plant failures

International Nuclear Information System (INIS)

Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.

1986-01-01

The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985

The program of reactors and nuclear power plants

International Nuclear Information System (INIS)

Calabrese, Carlos R.

2001-01-01

Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

Study on safety of a nuclear ship having an integral marine water reactor. Intelligent information database program concerned with thermal-hydraulic characteristics

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki; Kobayashi, Michiyuki; Murata, Hiroyuki; Aya, Izuo

2001-01-01

As a high economical marine reactor with sufficient safety functions, an integrated type marine water reactor has been considered most promising. At the National Maritime Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated/passive-safety type marine water reactor such as the flow boiling of a helical-coil type steam generator, natural circulation of primary water under a ship rolling motion and flashing-condensation oscillation phenomena in pool water has been conducted. This current study aims at making use of the safety analysis or evaluation of a future marine water reactor by developing an intelligent information database program concerned with the thermal-hydraulic characteristics of an integral/passive-safety reactor on the basis of the above-mentioned valuable experimental knowledge. Since the program was created as a Windows application using the Visual Basic, it is available to the public and can be easily installed in the operating system. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability limit for any helical-coil type once-through steam generator design. (2) analysis and comparison with the flow boiling data, (3) reference and graphic display of the experimental data, (4) indication of the knowledge information such as analysis method and results of the study. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor. (author)

Interactive nuclear plant analyzer for VVER-440 reactor

International Nuclear Information System (INIS)

Shier, W.; Horak, W.; Kennett, R.

1992-05-01

This document discusses an interactive nuclear plant analyzer (NPA) which has been developed for a VVER-440, Model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator

Evaluation of the trial design studies for an advanced marine reactor, (1)

International Nuclear Information System (INIS)

1988-03-01

The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

2001-07-01

Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

The Chernobyl reactor accident and its impact on the aquatic environment (marine and freshwater)

International Nuclear Information System (INIS)

Anon.

1987-01-01

The impact of the Chernobyl reactor accident of 1986 on the freshwater and marine environment of the British Isles is discussed. Particular reference is made to the monitoring of radionuclide concentrations in foodstuffs such as shellfish, molluscs, trout, watercress and laverbread and the possible restrictions on their consumption by the public. Attention is drawn to the effects of the accident on the marine environment in a wider context. (U.K.)

Marine ecosystem analysis for Kori nuclear power plant

International Nuclear Information System (INIS)

Lee, C.H.; Kim, Y.H.; Cho, T.S.

1980-01-01

The effect of both radioactive and thermal effluents discharged from the plant on aquatic ecosystem is one of the primary concerns in evaluating the environmental impact due to the operation of the nuclear power plant. Biological alteration of aquatic ecosystems may be resulted from radioactive effluents, thermal pollution and chemical releases. There is also another possible synergistic effect, that is, the combination of the above stresses, which may cause an impact severer than that of the sum of the individual impact. This report deals with species diversity and seasonal variations of those numbers of phytoplankton, marine algae and microorganisms, and distribution of radioactivity of marine organisms, as well as those data pertaining to sea water analysis. The present survey is designed to provide a partial baseline information for environmental impact assessment of Kori nuclear plant unit no. 1. (author)

Application of the system engineering approach for reactor plants design

International Nuclear Information System (INIS)

Sitskiy, S. B.

2010-01-01

The main activities planned for to be implemented are: developing a data model of the reactor plant plus integration with the information model of the plant (3D model + P & ID); reengineering of processes, developing of electronic documents; description of the equipment for information management of the reactor plant lifecycle – according ISO15926

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

Balance of plant improvements for future reactor projects

International Nuclear Information System (INIS)

Hollingshaus, H.

1987-01-01

Many studies have shown that improvements in portions of the plant other than the reactor systems can yield large cost savings during the design, construction, and operation of future reactor power plants. This portion is defined as the Balance of Plant which includes virtually everything except the equipment furnished by the Nuclear Steam Supply System manufacturer. It normally includes the erection of the entire plant including the NSSS. Cost of BOP equipment, engineering and construction work is therefore most of the cost of the plant. Improvements in the BOP have been identified that will substantially reduce nuclear plant cost and construction time while at the same time increasing availability and operability and improving safety. Improvements achieved through standardizatoin, simplification, three-dimensional (3D) computer-aided design, modular construction, innovative construction techniques, and applications for Artificial Intelligence Systems are described. (author)

Role of marine macroalgae in plant protection & improvement for sustainable agriculture technology

Directory of Open Access Journals (Sweden)

Seham M. Hamed

2018-03-01

Full Text Available Marine macroalgae are plant-like organisms with simple internal structures that generally live in coastal areas. They mainly include different communities of red, brown and green macroalgae. Marine macroalgae commonly occupy intertidal and sublittoral-to-littoral zones on rocks and other hard substrata. They are considered to be an excellent natural biosource in different aspects of agricultural fields. They have great proficiency in improving soil physical and chemical properties. Marine macroalgae are also characterized by producing a large array of biologically active biocidal substances against plant-infecting pathogens. Unfortunately, most available literatures on marine macroalgae and their derivatives mainly focused on their pharmaceutical applications but their potential utilization in sustainable agriculture development is still often regarded as a secondary goal. However, a relatively considerable dataset on marine macroalgae showed that they could play a major role in plant protection and improvement. This review summarizes different aspects of potential macroalgal applications in agriculture. Commercial production and exploitation of specific compounds with interesting biotechnological importance from marine macroalgae including microbicides, nematicides, insecticides, biofertilizers, biostimulators and soil conditioners are highlighted and discussed in detail. Bioactive compounds like fatty acids (in particular polyunsaturated fatty acids (PUFAs, proteins (amino acids, bioflavonoids, sulfated polysaccharides, carotenoids, polyphenols and carbohydrates are considered to have bactericidal, antiviral and fungicidal effects against some plant-infecting pathogens. These biocontrol agents provide multiple benefits and act as useful pointers for improving cultivation practices in diverse habitats. Marine macroalgae can be generally considered as promising multifunctional bioinoculants and ecofriendly environmental tools in recent trends

Marine-ecosystem analysis for the Kori nuclear power plant

International Nuclear Information System (INIS)

Lee, J.H.; Kim, Y.H.

1979-01-01

The effects of radioactive effluents and warm water discharged from the plant on aquatic ecosystem is one of the primary considerations in evaluating the impact due to the operation of the nuclear power plant. Biological alteration of aquatic ecosystems may be resulted from radioactive effluents, thermal pollution and chemical releases; there is also the possible synergistic effect, that is, the combination of the above stresses, which may cause an effect greater than that of the sum of the individual effects. This report deals with species diversity and seasonal vegetation of phytoplankton, marine algae and microorganisms, radioactive contamination of marine organisms, and lateral distribution of sea water temperature from discharge point. The present investigation is designed to provide a partial baseline information for environmental safety against Kori nuclear power plant. (author)

Marine macroscopic plants as biomass sources

Energy Technology Data Exchange (ETDEWEB)

North, W.J.

1979-01-01

Characteristics of marine plants, recent and current research, and studies at Woods Hole Oceanographic Institution and California Institute of Technology are reviewed. The latter program including laboratory and field studies on giant kelp is discussed. The use of deep ocean water and the nutrient requirements of giant kelp were studied. Test farm structure and problems are presented. (MHR)

Scenarios for potential radionuclide release from marine reactors dumped in the Kara Sea

International Nuclear Information System (INIS)

Lynn, N.; Mount, M.; Gussgard, K.

1995-01-01

The largest inventory of radioactive materials dumped in the Kara Sea by the former Soviet Union comes from the spent nuclear fuel (SNF) of seven marine reactors, the current (1994) inventory of which makes a total of approximately 4.7x10 15 Bq. In progressing its work for the International Arctic Seas Assessment Project, under the auspices of the International Atomic Energy Agency, the Source Term Working Group has analysed the Source Term and subsequently developed a number of model scenarios for the potential release patterns of radionuclides into the Kara Sea from the SNF and activated components dumped within the marine reactors.These models are based on the present and future conditions of the barrier materials and their configuration within the dumped objects. They account for progressive corrosion of the outer and inner steel barriers, breakdown of the organic fillers, and degradation and leaching from the SNFs. Annual release rates are predicted to four thousand years into the future. 5 refs., 1 fig., 1 tab

Nuclear reactor plant

International Nuclear Information System (INIS)

Schabert, H.P.; Laurer, E.

1977-01-01

The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

Potential of light water reactors for future nuclear power plants

International Nuclear Information System (INIS)

Gueldner, R.

2003-01-01

Energy consumption worldwide is going to increase further in the next few decades. Reliable supplies of electricity can be achieved only by centralized power plant structures. In this scenario, nuclear power plants are going to play a leading role as reliable and competitive plants, also under deregulated market conditions. Today, light water reactors have achieved a leading position, both technically and economically, contributing 85% to worldwide electricity generation in nuclear plants. They will continue to be a proven technology in power generation. In many countries, activities therefore are concentrated on extending the service life of plants beyond a period of forty years. New nuclear generating capacities are expected to be created and added from the end of this decade onward. Most of this capacity will be in light water reactors. The concepts of third-generation reactors will meet all economic and technical safety requirements of the 21st century and will offer considerable potential for further development. Probably some thirty years from now, fourth-generation nuclear power plants will be ready for commercial application. These plants will penetrate especially new sectors of the energy markets. Public acceptance of new nuclear power plants is not a matter of reactor lines, provided that safety requirements are met. The important issue is the management of radioactive waste. The construction of new nuclear power plants in Western Europe and North America mainly hinges on the ability to explain to the public that there is a need for new plants and that nuclear power is fundamental to assuring sustainable development. (orig.)

Radioecological studies of activation products released from a nuclear power plant into the marine environment

Energy Technology Data Exchange (ETDEWEB)

Nilsson, M.; Mattsson, S.; Holm, E.

1984-01-01

The Barseback nuclear power plant, located on the Oresund sound between Denmark and Sweden, consists of two boiling water reactors. The release of radionuclides, mainly activation products, is quite low during normal operation. During the summer, when annual overhaul and partial refuelling take place, the discharge is much higher. Samples of seaweeds and crustaceans collected along the coast were analyzed for radionuclides. Seaweeds (Fucus vesiculosus, F. Serratus, Ascophyllum nodosum and Cladophora glomerata) and crustaceans (Idothea and Gammarus) proved to be excellent bioindicators for radioactive corrosion products released from the nuclear power plant into the marine environment. These bioindicators have been used to map the spatial and temporal distribution of the released radioactivity. The activity has been followed up to 150 km from the power plant, and the decrease in activity concentration in the bioindicators with distance can be expressed by a power function. The variation with time of activity concentration reflects the amount of activity discharged from the power plant, with good resolution in time. The bioindicators exhibit different uptake patterns of the radionuclides detected. The crustacean Idothea showed variations in the Co/sup 60/ activity concentration between winter and summer. 9 references, 12 figures, 2 tables.

Radiological dose to man through the marine pathway from reactor operations at Humboldt Bay, California

International Nuclear Information System (INIS)

Noshkin, V.E.; Robison, W.L.; Harrison, F.L.

1976-01-01

Source-strength measurements and environmental samples taken at the Humboldt Bay Nuclear Reactor site near Eureka, California, since mid-1971 were used to evaluate the potential dose to man resulting from an aquatic release of radioactivity from the reactor. In this report, we provide an evaluation of individual and population dose through the marine pathways during 1972 and 1973 computed by the methods recommended by the U.S. Nuclear Regulatory Commission

In-plant considerations for optimal offsite response to reactor accidents

International Nuclear Information System (INIS)

Burke, R.P.; Heising, C.D.; Aldrich, D.C.

1982-11-01

Offsite response decision-making methods based on in-plant conditions are developed for use during severe reactor-accident situations. Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent-fuel storage pool from consideration for immediate offsite emergency response during accident situations. A simple plant information-management scheme is developed for use in offsite response decision-making. Detailed consequence calculations performed with the CRAC2 model are used to determine the appropriate timing of offsite-response implementation for a range of PWR accidents involving the reactor core. In-plant decision criteria for offsite-response implementation are defined. The definition of decision criteria is based on consideration of core-accident physical processes, in-plant accident monitoring information, and results of consequence calculations performed to determine the effectiveness of various public-protective measures. The benefits and negative aspects of the proposed response-implementation criteria are detailed

Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

International Nuclear Information System (INIS)

1979-11-01

This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

Field Keys to Common Hawaiian Marine Animals and Plants.

Science.gov (United States)

Hawaii State Dept. of Education, Honolulu. Office of Instructional Services.

Presented are keys for identifying common Hawaiian marine algae, beach plants, reef corals, sea urchins, tidepool fishes, and sea cucumbers. Nearly all species considered can be distinguished by characteristics visible to the naked eye. Line drawings illustrate most plants and animals included, and a list of suggested readings follows each…

Advanced liquid metal reactor plant control system

International Nuclear Information System (INIS)

Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

1993-01-01

The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

EL-Kafas, A.E.A.E.

1996-01-01

the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

Liquid Metal Fast Breeder Reactor plant maintenance and equipment design

International Nuclear Information System (INIS)

Swannack, D.L.

1982-01-01

This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment

Design study of plant system for the fusion experimental reactor (FER)

International Nuclear Information System (INIS)

Iida, Hiromasa; Kuroda, Hideo; Yamada, Masao; Suzuki, Tatsushi; Honda, Tsutomu; Ohmura, Hiroshi; Itoh, Shinichi.

1986-11-01

This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

Integrated gasification gas combined cycle plant with membrane reactors: Technological and economical analysis

International Nuclear Information System (INIS)

Amelio, Mario; Morrone, Pietropaolo; Gallucci, Fausto; Basile, Angelo

2007-01-01

In the present work, the capture and storage of carbon dioxide from the fossil fuel power plant have been considered. The main objective was to analyze the thermodynamic performances and the technological aspects of two integrated gasification gas combined cycle plants (IGCC), as well as to give a forecast of the investment costs for the plants and the resulting energy consumptions. The first plant considered is an IGCC* plant (integrated gasification gas combined cycle plant with traditional shift reactors) characterized by the traditional water gas shift reactors and a CO 2 physical adsorption system followed by the power section. The second one is an IGCC M plant (integrated gasification gas combined cycle plant with membrane reactor) where the coal thermal input is the same as the first one, but the traditional shift reactors and the physical adsorption unit are replaced by catalytic palladium membrane reactors (CMR). In the present work, a mono-dimensional computational model of the membrane reactor was proposed to simulate and evaluate the capability of the IGCC M plant to capture carbon dioxide. The energetic performances, efficiency and net power of the IGCC* and IGCC M plants were, thus, compared, assuming as standard a traditional IGCC plant without carbon dioxide capture. The economical aspects of the three plants were compared through an economical analysis. Since the IGCC* and IGCC M plants have additional costs related to the capture and disposal of the carbon dioxide, a Carbon Tax (adopted in some countries like Sweden) proportional to the number of kilograms of carbon dioxide released in the environment was assumed. According to the economical analysis, the IGCC M plant proved to be more convenient than the IGCC* one

The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1979-01-01

The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

Energy Technology Data Exchange (ETDEWEB)

1979-11-01

This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies.

Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

International Nuclear Information System (INIS)

1979-11-01

This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies

Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

International Nuclear Information System (INIS)

1986-09-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

Ensuring the operational safety of nuclear power plants with WWER reactors

International Nuclear Information System (INIS)

Shasharin, G.A.; Veretennikov, G.A.; Abagyan, A.A.; Lesnoj, S.A.

1984-01-01

At the start of 1983, 27 nuclear power producing units with reactor facilities of the WWER type were in operation in the Soviet Union and other countries. In 1982 the average load factor for nuclear power plants with WWER reactors was 73 per cent. There was not a single nuclear accident or even damage with any significant radiation consequences in the WWER reactors during the entire period of their operation. The most modern nuclear power plants with WWER-440 and WWER-1000 reactors meet all present-day international requirements. Safe operation of the plants is achieved by a variety of measures, the most important of which include: procedures for increasing the reliability of plant equipment and systems; ensuring exact compliance with plant operating instructions; ensuring reliable operation of plant safety systems; action directed towards maintaining the skills of plant personnel at a level adequate to ensure the taking of proper action during transient processes and accident situations. The paper discusses concrete steps for ensuring safe nuclear power plant operation along these lines. In particular, measures such as the following are described: the use of a system for collecting and processing information on equipment failures and defects; the development and introduction of methods of early defect diagnosis; the performance of complex testing of safety systems; the training of highly skilled personnel for nuclear power plants at educational combines and at teaching and training centres making use of simulators; arranging accident-prevention training and special instruction for personnel. (author)

The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

International Nuclear Information System (INIS)

Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

2012-01-01

This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

Effects of ship casualties on reactor safety and marine reactor design

International Nuclear Information System (INIS)

Agena, H.H.

1978-01-01

Ship casualties such as collision, grounding, fire and explosions, leakage and partial flooding may severely impair the safety of the nuclear reactor plant and result in nuclear hazards to the crew and the environment. Engineered safeguards are being discussed for protection against such consequences: manoeuvrability, structural (passive) collision and fire protection, protection against external fires and pressure waves, after heat transmission in case of a casualty, and after heat transmission out off the sunk wreck. Existing requirements will be discussed, shortly

Nitrogen and phosphorus treatment of marine wastewater by a laboratory-scale sequencing batch reactor with eco-friendly marine high-efficiency sediment.

Science.gov (United States)

Cho, Seonghyeon; Kim, Jinsoo; Kim, Sungchul; Lee, Sang-Seob

2017-06-22

We screened and identified a NH 3 -N-removing bacterial strain, Bacillus sp. KGN1, and a [Formula: see text] removing strain, Vibrio sp. KGP1, from 960 indigenous marine isolates from seawater and marine sediment from Tongyeong, South Korea. We developed eco-friendly high-efficiency marine sludge (eco-HEMS), and inoculated these marine bacterial strains into the marine sediment. A laboratory-scale sequencing batch reactor (SBR) system using the eco-HEMS for marine wastewater from land-based fish farms improved the treatment performance as indicated by 88.2% removal efficiency (RE) of total nitrogen (initial: 5.6 mg/L) and 90.6% RE of total phosphorus (initial: 1.2 mg/L) under the optimal operation conditions (food and microorganism (F/M) ratio, 0.35 g SCOD Cr /g mixed liquor volatile suspended solids (MLVSS)·d; dissolved oxygen (DO) 1.0 ± 0.2 mg/L; hydraulic retention time (HRT), 6.6 h; solids retention time (SRT), 12 d). The following kinetic parameters were obtained: cell yield (Y), 0.29 g MLVSS/g SCOD Cr ; specific growth rate (µ), 0.06 d -1 ; specific nitrification rate (SNR), 0.49 mg NH 3 -N/g MLVSS·h; specific denitrification rate (SDNR), 0.005 mg [Formula: see text]/g MLVSS·h; specific phosphorus uptake rate (SPUR), 0.12 mg [Formula: see text]/g MLVSS·h. The nitrogen- and phosphorus-removing bacterial strains comprised 18.4% of distribution rate in the microbial community of eco-HEMS under the optimal operation conditions. Therefore, eco-HEMS effectively removed nitrogen and phosphorus from highly saline marine wastewater from land-based fish farms with improving SNR, SDNR, and SPUR values in more diverse microbial communities. DO: dissolved oxygen; Eco-HEMS: eco-friendly high efficiency marine sludge; F/M: food and microorganism ratio; HRT: hydraulic retention time; ML(V)SS: mixed liquor (volatile) suspended solids; NCBI: National Center for Biotechnology Information; ND: not determined; qPCR: quantitative real-time polymerase

Review of advanced reactor transient analysis capabilities and applications for Savannah River Plant reactors

International Nuclear Information System (INIS)

Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.

1977-01-01

GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations

Remote repairs and refurbishment of reactor internal structures of magnox plant

International Nuclear Information System (INIS)

Barnes, S.A.; Kelly, D.E.

1992-01-01

The original designers of the UK Magnox reactor plant made provision for the then perceived time dependent processes that could have influenced the operational life of the plant. Changes in graphite properties with irradiation, particularly dimensional change, were well understood and in-core samples were provided for subsequent laboratory examination to monitor the processes throughout plant life. The tendency towards embrittlement with irradiation of the steel of the reactor pressure vessels was also acknowledged and again in-core samples were provided for monitoring changes in materials properties in-service and thus provide data in support of structural analyses to sustain the reactor safety cases. (author)

Studies on plant dynamics of sodium-cooled fast breeder reactors - verification of a plant model

International Nuclear Information System (INIS)

Schubert, B.

1988-01-01

For the analysis of sodium-cooled FBR safety and dynamics theoretical models are used, which have to be verified. In this report the verification of the plant model SSC-L is conducted by the comparison of calculated data with measurements of the experimental reactors KNK II and RAPSODIE. For this the plant model is extended and adapted. In general only small differences between calculated and measured data are recognized. The results are used to improve and complete the plant model. The extensions of the plant model applicability are used for the calculation of a loss of heat sink transient with reactor scram, considering pipes as passive heat sinks. (orig./HP) With 69 figs., 10 tabs [de

Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

International Nuclear Information System (INIS)

1984-06-01

The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

Inventory and source term evaluation of Russian nuclear power plants for marine applications

International Nuclear Information System (INIS)

Reistad, O.; Oelgaard, P.L.

2006-04-01

This report discusses inventory and source term properties in regard to operation and possible releases due to accidents from Russian marine reactor systems. The first part of the report discusses relevant accidents on the basis of both Russian and western sources. The overview shows that certain vessels were much more accident prone compared to others, in addition, there have been a noteworthy reduction in accidents the last two decades. However, during the last years new types of incidents, such as collisions, has occurred more frequently. The second part of the study considers in detail the most important factors for the source term; reactor operational characteristics and the radionuclide inventory. While Russian icebreakers has been operated on a similar basis as commercial power plants, the submarines has different power cyclograms which results in considerable lower values for fission product inventory. Theoretical values for radionuclide inventory are compared with computed results using the modelling tool HELIOS. Regarding inventory of transuranic elements, the results of the calculations are discussed in detail for selected vessels. Criticality accidents, loss-of-cooling accidents and sinking accidents are considered, bases on actual experiences with these types of accident and on theoretical considerations, and source terms for these accidents are discussed in the last chapter. (au)

Inventory and source term evaluation of Russian nuclear power plants for marine applications

Energy Technology Data Exchange (ETDEWEB)

Reistad, O. [Norwegian Radiation Protection Authority (Norway); Oelgaard, P.L. [Risoe National Lab. (Denmark)

2006-04-15

This report discusses inventory and source term properties in regard to operation and possible releases due to accidents from Russian marine reactor systems. The first part of the report discusses relevant accidents on the basis of both Russian and western sources. The overview shows that certain vessels were much more accident prone compared to others, in addition, there have been a noteworthy reduction in accidents the last two decades. However, during the last years new types of incidents, such as collisions, has occurred more frequently. The second part of the study considers in detail the most important factors for the source term; reactor operational characteristics and the radionuclide inventory. While Russian icebreakers has been operated on a similar basis as commercial power plants, the submarines has different power cyclograms which results in considerable lower values for fission product inventory. Theoretical values for radionuclide inventory are compared with computed results using the modelling tool HELIOS. Regarding inventory of transuranic elements, the results of the calculations are discussed in detail for selected vessels. Criticality accidents, loss-of-cooling accidents and sinking accidents are considered, bases on actual experiences with these types of accident and on theoretical considerations, and source terms for these accidents are discussed in the last chapter. (au)

Seawater desalination plant using nuclear heating reactor coupled with MED process

Institute of Scientific and Technical Information of China (English)

无

2000-01-01

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. This seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10～200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m3/d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented.

Seawater desalination plant using nuclear heating reactor coupled with MED process

International Nuclear Information System (INIS)

Wu Shaorong; Dong Duo; Zhang Dafang; Wang Xiuzhen

2000-01-01

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. this seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. the intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10-200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m 3 /d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented

Safety analysis of an expert reactor protection system in nuclear power plants

International Nuclear Information System (INIS)

El-Kafas, A.A.

1997-01-01

The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

Dynamic simulation of a sodium-cooled fast reactor power plant

Energy Technology Data Exchange (ETDEWEB)

Shinaishin, M.A.M.

1976-08-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

Dynamic simulation of a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Shinaishin, M.A.M.

1976-08-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant

Analysis of reactor trips originating in balance of plant systems

International Nuclear Information System (INIS)

Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W.

1990-09-01

This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs

STEADY-SHIP: a computer code for three-dimensional nuclear and thermal-hydraulic analyses of marine reactors

International Nuclear Information System (INIS)

Itagaki, Masafumi; Naito, Yoshitaka; Tokuno, Yukio; Matsui, Yasushi.

1988-01-01

A code STEADY-SHIP has been developed to calculate three-dimensional distributions of neutron flux, power and coolant temperature in the reactor core of the nuclear ship MUTSU. The code consists of two parts, that is, a few-group three-dimensional neutron diffusion module DIFFUSION-SHIP and a thermal-hydraulic module HYDRO-SHIP: In the DIFFUSION-SHIP the leakage iteration method is used for solving the three-dimensional neutron diffusion equation with small computer core memory and short computing time; The HYDRO-SHIP performs the general thermal-hydraulic calculation for evaluating feedbacks required in the neutronic calculation by the DIFFUSION-SHIP. The macroscopic nuclear constants are generated by a module CROSS-SHIP as functions of xenon poison, fuel temperature, moderator temperature and moderator density. A module LOCAL-FINE has the capability of computing a detailed rod power distribution for each local node in the core, using the boundary conditions on the surface of the node which were supplied by the STEADY-SHIP whole-core calculation. The applicability of this code to marine reactors has been demonstrated by comparing the computed results with the data measured during the MUTSU land-loaded core critical experiments and with the data obtained during the hot-zero-power tests performed for the actual MUTSU plant. (author)

Coupling of RO-MSF hybrid desalination plants with nuclear reactors

International Nuclear Information System (INIS)

Al-Sulaiman, Khalil; Al-Mutaz, Ibrahim S.

1999-01-01

Full text.Reverse osmosis (RO) and multistage flash (MSF) desalination are the most widely commercial available processes. MSF utilizes stream in the brine heater as a primary source of energy. RO is derived mainly by electricity that pumps the feed water against the mambranes. Steam and electricity and be produced easily by nuclear reactors. Nuclear reactors may be coupled with deslination plants (MSF, RO or combined (hybrid) RO/MSF configuration). This integrated plant will be capable of producing power and water at reasonable cost. The capital and operating cost will be reduced and the excess power can be efficiently utilized. Maintenance and operating cost will drop significantly. In this paper, a techno-economic study of hybrid reverses osmosis /multistage flash desalination will be carried. The proposed configuration (hybrid RO/MSF) coupled with nuclear reactor is considered the most appropriate candidate system for the application of dual-purpose nuclear desalination plants. the design parameters for such a desalination hybrid system will be the applied pressure and recovery for reverse osmosis plant and the number of stages and the heat transfer areas for multistage flash plant

Power unit with GT-MHR reactor plant for electricity production and district heating

International Nuclear Information System (INIS)

Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

2000-01-01

Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

Model-Based Analysis and Efficient Operation of a Glucose Isomerization Reactor Plant

DEFF Research Database (Denmark)

Papadakis, Emmanouil; Madsen, Ulrich; Pedersen, Sven

2015-01-01

efficiency. The objective of this study is the application of the developed framework on an industrial case study of a glucose isomerization (GI) reactor plant that is part of a corn refinery, with the objective to improve the productivity of the process. Therefore, a multi-scale reactor model...... is developedfor use as a building block for the GI reactor plant simulation. An optimal operation strategy is proposed on the basis of the simulation results...

Plant accident dynamics of high-temperature reactors with direct gas turbine cycle

International Nuclear Information System (INIS)

Waloch, M.L.

1977-01-01

In the paper submitted, a one-dimensional accident simulation model for high-temperature reactors with direct-cycle gas turbine (single-cycle facilities) is described. The paper assesses the sudden failure of a gas duct caused by the double-ended break of one out of several parallel pipes before and behind the reactor for a non-integrated plant, leading to major loads in the reactor region, as well as the complete loss of vanes of the compressor for an integrated plant. The results of the calculations show especially high loads for the break of a hot-gas pipe immediately behind the flow restrictors of the reactor outlet, because of prolonged effects of pressure gradients in the reactor region and the maximum core differential pressure. A plant accident dynamics calculation therefore allows to find a compromise between the requirements of stable compressor operation, on the one hand, and small loads in the reactor in the course of an accident, on the other, by establishing in a co-ordinated manner the narrowing ratio of the flow restrictors. (GL) [de

High-rate nitrogen removal from waste brine by marine anammox bacteria in a pilot-scale UASB reactor.

Science.gov (United States)

Yokota, Nobuyuki; Watanabe, Yasutsugu; Tokutomi, Takaaki; Kiyokawa, Tomohiro; Hori, Tomoyuki; Ikeda, Daisuke; Song, Kang; Hosomi, Masaaki; Terada, Akihiko

2018-02-01

The goal of this study was to develop a startup strategy for a high-rate anaerobic ammonium oxidation (anammox) reactor to treat waste brine with high concentrations of ammonium from a natural gas plant. An upflow anaerobic sludge blanket (UASB) anammox reactor with an effective volume of 294 L was fed continuously with waste brine with a salinity of 3% and a NH 4 + concentration of 180 mg-N/L, as well as a NaNO 2 solution. By inoculating a methanogenic granular biomass as a biomass carrier, the reactor attained the maximum volumetric nitrogen removal rate (NRR) of 10.7 kg-N/m 3 /day on day 209, which was 1.7 times higher than the highest reported NRR for wastewater of comparable salinity. High-throughput sequencing of 16S rRNA gene amplicons revealed that Candidatus Scalindua wagneri was enriched successfully in granules in the UASB, and it replaced Methanosaeta and became dominant in the granule. The inhibitory effect of NO 2 - on the anammox reaction in the granules was assessed by a 15 N tracer method, and the results showed that anammox activity was maintained at 60% after exposure to 300 mg-N/L of NO 2 - for 24 h. Compared with previous studies of the susceptibilities of Candidatus Brocadia and Candidatus Kuenenia to NO 2 - , the enriched marine anammox bacteria were proven to have comparable or even higher tolerances for high NO 2 - concentrations after a long exposure.

Tritium instrumentation for a fusion reactor power plant

International Nuclear Information System (INIS)

Shank, K.E.; Easterly, C.E.

1976-09-01

A review of tritium instrumentation is presented. This includes a discussion of currently available in-plant instrumentation and methods required for sampling stacks, monitoring process streams and reactor coolants, analyzing occupational work areas for air and surface contamination, and personnel monitoring. Off-site instrumentation and collection techniques are also presented. Conclusions are made concerning the adequacy of existing instrumentation in relation to the monitoring needs for fusion reactors

Interactive nuclear plant analyzer for the VVER-440 reactor

International Nuclear Information System (INIS)

Shier, W.; Kennett, R.

1993-01-01

An interactive nuclear plant analyzer (NPA) has been developed for a VVER-440 model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator. Results of the interactive calculation can be through the user-defined, digital display of various plant parameters and through color changes that reflect changes in primary system fluid temperatures, fuel and clad temperatures, and the temperatures of other metal structures. In addition, changes in the status of various components and system can be initiated and/or displayed both numerically and graphically on the mask

Enrichment using an up-flow column reactor and community structure of marine anammox bacteria from coastal sediment.

Science.gov (United States)

Kindaichi, Tomonori; Awata, Takanori; Suzuki, Yuji; Tanabe, Katsuichiro; Hatamoto, Masashi; Ozaki, Noriatsu; Ohashi, Akiyoshi

2011-01-01

We established an enrichment culture of marine anaerobic ammonium oxidation (anammox) bacteria using an up-flow column reactor fed with artificial sea water supplemented with nitrogen and minerals and inoculated with coastal surface sediment collected from Hiroshima Bay. After 2 months of reactor operation, simultaneous removal of NH(4)(+) and NO(2)(-) was observed, suggesting that an anammox reaction was proceeding. A total nitrogen removal rate of 2.17 g-N L(-1) day(-1) was attained on day 594 while the nitrogen loading rate was 3.33 g-N L(-1) day(-1). Phylogenetic analysis revealed that at least two dominant "Candidatus Scalindua" species were present in this reactor. Moreover, many uncultured bacteria and archaea, including candidate division or ammonia-oxidizing archaea, were present. Fluorescence in situ hybridization (FISH) revealed that anammox bacteria accounted for 85.5 ± 4.5% of the total bacteria at day 393. We also designed two oligonucleotide probes specific to each dominant "Candidatus Scalindua" species. A simultaneous FISH analysis using both probes showed that two different "Candidatus Scalindua" species were clearly recognizable and coexisted during reactor operation, although there was some variation in their abundance. The marine anammox bacteria enriched in this study have potential applications to the treatment of industrial wastewater containing high levels of ammonium and salt.

Angra nuclear plant - environmental control program

International Nuclear Information System (INIS)

Kircher, E.; Cruz, E.S. da

1989-01-01

The pre-operational studies, that were elaborated before the beginning of Angra I Power Plant operation, are described in particular the environmental radiological safety area till the fuel loading in the core reactor. Several aspects are included, as socio-economic survey, seismological analysis, Meteorological Program, marine biology, water cooling system, exposure measures of natural radiation, marine sediments characterization in the effluent dispersion area and Environmental Radiological Monitoring Program. The main environmental programs developed for the operational phase of the Angra I Plant are also presented, citing some considerations about the Meteorological Program, Marine Biology Control Program, Temperature and Chlorine Control in Piraquara de Fora Bay, Radiological Environmental Monitoring Program, Sanitary Effluent Control Program and Radiological Emergency Program. (C.G.C.). 2 refs

Plant volatiles in extreme terrestrial and marine environments.

Science.gov (United States)

Rinnan, Riikka; Steinke, Michael; McGenity, Terry; Loreto, Francesco

2014-08-01

This review summarizes the current understanding on plant and algal volatile organic compound (VOC) production and emission in extreme environments, where temperature, water availability, salinity or other environmental factors pose stress on vegetation. Here, the extreme environments include terrestrial systems, such as arctic tundra, deserts, CO₂ springs and wetlands, and marine systems such as sea ice, tidal rock pools and hypersaline environments, with mangroves and salt marshes at the land-sea interface. The emission potentials at fixed temperature and light level or actual emission rates for phototrophs in extreme environments are frequently higher than for organisms from less stressful environments. For example, plants from the arctic tundra appear to have higher emission potentials for isoprenoids than temperate species, and hypersaline marine habitats contribute to global dimethyl sulphide (DMS) emissions in significant amounts. DMS emissions are more widespread than previously considered, for example, in salt marshes and some desert plants. The reason for widespread VOC, especially isoprenoid, emissions from different extreme environments deserves further attention, as these compounds may have important roles in stress resistance and adaptation to extremes. Climate warming is likely to significantly increase VOC emissions from extreme environments both by direct effects on VOC production and volatility, and indirectly by altering the composition of the vegetation. © 2014 John Wiley & Sons Ltd.

Capital cost: gas cooled fast reactor plant

International Nuclear Information System (INIS)

1977-09-01

The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design

Chinese nuclear heating test reactor and demonstration plant

International Nuclear Information System (INIS)

Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

1992-01-01

In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

A capital cost reduction study on the fast breeder reactor plant

International Nuclear Information System (INIS)

Taniyama, H.; Kamei, M.; Moriyama, M.

1991-01-01

A capital cost reduction study has been performed for large fast breeder reactor designs. The primary objective of this study is to show a trend of capital cost reduction between FBR plants at the prototype stage, the demonstration stage, and the future commercialization stage. For the FBR plant at the demonstration stage a construction cost comparison with a light water reactor has also been performed, and the target cost of FBR of below 1.5 times that of the light water reactor cost was achieved. To extend the capital cost reduction study, a feasibility study was made to achieve a capital cost of an FBR less than that of a light water reactor. The recommended design is shown as a future commercialization FBR design concept. (author)

Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Souza Mendes, J.E. de

1987-01-01

Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

Plant with nuclear reactor, in particular a thermal reactor

International Nuclear Information System (INIS)

Straub, H.

1988-01-01

The reactor core of the plant has tubular and vertically movable control rods moved by a flow of coolant under pressure. Each control rod surrounds a similarly tubular guide rod, stationary relative to the reactor core, leaving an annular slot-like space therebetween. The inside of each guide rod forms a first pressure chamber supplied with the coolant under pressure. The upper end of each control rod is closed and has a vertical shaft that extends into the inside of the guide rod and forms therewith a second annular slot-like space. At least one first restriction is provided in the first annular slot-like space and at least one second restriction is provided in the second annular slot-like space. A second pressure chamber is formed between both restrictions. The coolant supplied to the guide rod thus returns to the pressure vessel surrounding the reactor core through the second annular slot-like space, the second pressure chamber and the first annular slot-like space. Controlling means are provided, with which pressure thrusts can be generated if necessary in the coolant within the first pressure chamber. (author) 5 refs., 10 figs

Nuclear reactor plants and control systems therefor

International Nuclear Information System (INIS)

de Boer, G.A.; de Hex, M.

1976-01-01

A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

Clinch River Breeder Reactor Plant Project: construction schedule

International Nuclear Information System (INIS)

Purcell, W.J.; Martin, E.M.; Shivley, J.M.

1982-01-01

The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

Advanced marine reactor MRX and application to nuclear barge supplying electricity and heat

International Nuclear Information System (INIS)

Ishida, Toshihisa; Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Fukuhara, Yoshifumi; Ochiai, Masa-aki

2000-01-01

The basic design concept of an advanced marine reactor MRX has been established with adoption of several new technologies. The MRX is an integral-type PWR with 100 MWt aimed basically for use of ship propulsion. Adoption of a water-filled containment together with the integral type reactor makes the reactor light-weight and compact greatly. A engineered safety system is a simplified passive system, function of which is confirmed by the safety analysis. The MRX can be applied to an energy supply system of electricity and heat co-generation by installing it on a barge. Concept of a nuclear barge with the MRX of 334 MWt output is presented for use of supplying electricity, fresh water and hot water. Combined system of electric generation and desalination with the RO process can deliver variable output of electricity and fresh water according a demand. Latent heat of the exhausted steam from the turbine can be used effectively to raise the temperature of cold water as heat supply. (author)

Reliability of reactor plant water cleanup pumps

International Nuclear Information System (INIS)

Pearson, J.L.

1979-01-01

Carolina Power and Light Company's Brunswick 2 nuclear plant experienced a high reactor water cleanup pump-failure rate until inlet temperature and flow were reduced and mechanical modifications were implemented. Failures have been zero for about one year, and water cleanup efficiency has increased

Increase of efficiency of plant materials heat treatment in tubular reactors

Directory of Open Access Journals (Sweden)

A. V. Golubkovich

2016-01-01

Full Text Available In agriculture products of pyrolysis of plant materials in the form of waste of the main production can be applied as a source of heat and electric power. Besides, their use prevents ecological pollution of the soil and the atmosphere. Pyrolysis plants can be used for work with tubular reactors anywhere. Due to them farmers can dry grain, using waste heat of diesel generators, heatgenerators, boiler plants and receiving thus gaseous products, liquid and firm fractions. A technology based on cyclic and continuous plant mass movement by a piston in a pipe from a loading site to a place of unloading of a firm phase consistently through cameras of drying, pyrolysis, condensation of gaseous products. Exhaust furnace gases with a temperature up to 600 degrees Celsius are given countercurrent material movement from a power equipment. The gaseous, liquid and firm products from the pyrolysis camera are used for heat and electric power generation. Calculation of parameters of subdrying and pyrolysis cameras is necessary for effective and steady operation of the tubular reactor. The authors determined the speed of raw materials movement, and also duration of drying and pyrolysis in working chambers. An analysis of a simplified mathematical model of process was confirmed with results of experiments. Models of heat treatment of wet plant materials in tubular reactors are worked out on a basis of equality of speeds of material movement in the reactor and distribution of a temperature front in material on radius. The authors defined estimated characteristic for determination of tubular reactor productivity and size of heat, required for drying and pyrolysis.

At-reactor storage of spent fuel for life-of-plant

International Nuclear Information System (INIS)

Fuierer, A.A.

1990-01-01

The management of commercial spent fuel is a fairly broad topic beginning with the discharge from a reactor, its storage on-site, its transport from the reactor site to a U.S. Department of Energy (DOE) facility, and its ultimate disposal in a geologic repository. This paper discusses spent-fuel management in the at-reactor phase. There are two basic methods for at-reactor storage of spent fuel. The first is wet storage in a pool, and the second is dry storage external to the plant in some form of cask or vault. Spent-fuel consolidation will impact the utility and the DOE waste system. Some of these impacts have a positive effect and some have a negative effect, and each will vary somewhat for each utility. Spent-fuel consolidation and life-of-plant storage will be an increased burden to utilities but will likely result in significant cost savings to the overall waste management system and by proper integration can result in significant institutional benefits

Steam turbine chemistry in light water reactor plants

International Nuclear Information System (INIS)

Svoboda, Robert; Haertel, Klaus

2008-01-01

Steam turbines in boiling water reactor (BWR) and pressurized water reactor (PWR) power plants of various manufacturers have been affected by corrosion fatigue and stress corrosion cracking. Steam chemistry has not been a prime focus for related research because the water in nuclear steam generating systems is considered to be of high purity. Steam turbine chemistry however addresses more the problems encountered in fossil fired power plants on all volatile treatment, where corrosive environments can be formed in zones where wet steam is re-evaporated and dries out, or in the phase transition zone, where superheated steam starts to condense in the low-pressure (LP) turbine. In BWR plants the situation is aggravated by the fact that no alkalizing agents are used in the cycle, thus making any anionic impurity immediately acidic. This is illustrated by case studies of pitting corrosion of a 12 % Cr steel gland seal and of flow-oriented corrosion attack on LP turbine blades in the phase transition zone. In PWR plants, volatile alkalizing agents are used that provide some buffering of acidic impurities, but they also produce anionic decomposition products. (orig.)

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

International Nuclear Information System (INIS)

1986-06-01

The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant

High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-06-15

The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant.

Prestressed reactor vessel for nuclear power plants

International Nuclear Information System (INIS)

Schoening, J.; Schwiers, H.G.

1982-01-01

With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

Main gas circulator for VG-400 reactor plant

International Nuclear Information System (INIS)

Mitenkov, F.M.; Kostin, V.I.; Novinskij, E.G.; Kuropatov, A.I.; Protsenko, A.N.; Smirnov, V.P.; Stolyarevskij, A.Ya.

1988-01-01

Principle parameters and operating conditions of the main gas circulator (MGC) in VG-400 reactor plant are presented. Brief MGC design description and experimental work scope are given. (author). 4 refs, 4 figs, 1 tab

L-Reactor operation, Savannah River Plant: environmental assessment

International Nuclear Information System (INIS)

1982-08-01

The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

Overall plant concept for a tank-type fast reactor

International Nuclear Information System (INIS)

Yamaki, Hideo; Davies, S.M.; Goodman, L.

1984-01-01

Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers

Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

International Nuclear Information System (INIS)

Chang H. Oh; Eung Soo Kim; Steven Sherman

2008-01-01

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood

Plant Control Concept for the Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Eui Kwang; Kim, S. O.

2010-12-01

A power plant is designed for incorporation into a utility's grid system and follows the load demand through the steam generator, intermediate heat exchanger(IHX), from the nuclear core. During the load-following transients, various plant parameters must be controlled to protect the reactor core and other components in the plant. The purpose of this report is to review design considerations to establish SFR plant control and to design plant control concepts. The governing equations and solution procedure of the computer code to calculate plant temperature conditions during the part-load operation was reviewed and 4 types of plant operation concepts were designed, and the results of the calculations were compared

Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-08-01

This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

International Nuclear Information System (INIS)

1995-08-01

This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory

Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

Energy Technology Data Exchange (ETDEWEB)

Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens, E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas

2015-07-01

The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

International Nuclear Information System (INIS)

Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens

2015-01-01

The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

The use of marine aquaculture solid waste for nursery production of the salt marsh plants Spartina alterniflora and Juncus roemerianus

Directory of Open Access Journals (Sweden)

H.M. Joesting

2016-05-01

Full Text Available Recent technological advances in marine shrimp and finfish aquaculture alleviate many of the environmental risks associated with traditional aquaculture, but challenges remain in cost-effective waste management. Liquid effluent from freshwater aquaculture systems has been shown to be effective in agricultural crop production (i.e., aquaponics, but few studies have explored the potential for reuse of marine aquaculture effluent, particularly the solid fraction. The purpose of this study was to investigate the use of marine aquaculture solid waste as a nutrient source for the nursery production of two salt tolerant plants commonly used in coastal salt marsh restoration, Spartina alterniflora (smooth cordgrass and Juncus roemerianus (black needlerush. Specifically, measurements of plant biomass and tissue nitrogen and phosphorus allocation were compared between plants fertilized with dried shrimp biofloc solids and unfertilized controls, as well as between plants fertilized with dried fish solids and unfertilized controls. In both experiments, S. alterniflora plants fertilized with marine aquaculture solids showed few significant differences from unfertilized controls, whereas fertilized J. roemerianus plants had significantly greater biomass and absorbed and incorporated more nutrients in plant tissue compared to unfertilized controls. These results suggest that J. roemerianus may be a suitable plant species for the remediation of marine aquaculture solid waste. Keywords: Marine aquaculture, Salt marsh plants, Solid waste, Phytoremediation

Training nuclear power plant personnel on SR-O reactor

International Nuclear Information System (INIS)

Cerny, K.; Boucek, F.; Kveton, M.; Prokopec, Z.; Fleischhans, J.

1983-01-01

The SR-O reactor is an experimental pool-type reactor with a maximum output of 1 MW and maximum thermal neutron flux density of 5.3x10 13 m -2 s -1 . The reactor is described in detail and its specifications are given. The protection and control systems of the reactor permit both manual and automatic operation. The reactor is used for training courses for nuclear power plant operators and for post-graduate study courses for other specialists. Intensive courses for 4 to 6 persons take 15 to 20 days. The course is adjusted to the results of introductory theoretical tests. An optimal teaching method has been developed based on the flowchart algorithmic method, dividing activities into operations (manipulations with controls, issuing commands, making records, etc.) and decision making (information reception and processing). (M.D.)

Plant-scale anodic dissolution of unirradiated N-Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1995-01-01

Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

Marine monitoring surveys for desalination plants-A critical review

KAUST Repository

Lattemann, Sabine

2013-01-01

Environmental impact assessment (EIA) studies are standard practice and a regulatory requirement for most new desalination projects today. However, most of the EIA studies are limited to predictive information; that is, they gather information on the project and the project\\'s environment before project implementation to make predictions about likely impacts. The EIAs may involve comprehensive studies, such as field monitoring, laboratory toxicity testing, and modeling studies. Consequently, the"surprising paucity of useful experimental data, either from laboratory tests or from field monitoring studies", which was observed by the US National Research Council in 2008, has been gradually decreasing. However, there is still a long-term research need on the site-specific effects of desalination plants after project commissioning has taken place. A main challenge of field research is the adequate design of the monitoring studies, which have to adequately distinguish the effects of the desalination project from natural processes over long periods of time. The existing monitoring studies have so far used a wide range of approaches and methods to investigate the environmental impacts of desalination plant discharges. Shortfalls are often that they are limited in scope, short-term, or localized. In essence, many studies fall short of recognizing the potentially synergetic effects of the single waste components of the discharges on marine organisms and the complexity of the potential responses by the ecosystem. While the possible risk of damage arising from the concentrate discharge to the marine environment in close proximity to the outfall is at hand, no conclusive evidence can yet be provided concerning the long-term impacts of desalination plant discharges, let alone the cumulative impacts on certain sea areas. This paper conducts a critical review of existing monitoring programs for desalination plants. Shortcomings of current practices are identified and relevant

Study on dual plant concept for the next generation boiling water reactors

International Nuclear Information System (INIS)

Sato, Takashi; Oikawa, Hirohide

1999-01-01

The paper presents the study results on the basic concept of dual BWRs. For the convenience, we call the concept here as Trial Study on BWR dual concept (TSBWR dual). The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The TSBWR dual is a plant concept of dual BWRs contained in a same secondary containment building. The plant output is from 2 x l,350 MWe up to 2 x 1,700 MWe. This concept is mainly aiming at safety improvement and cost savings of the next generation BWRs. The TSBWR dual has two RPVs and two dry wells (DW). It has, however, only one wet well (WW) and only one R/B. The WW and the R/B are shared by the dual reactors. The operating floor is also shared by the two reactors. The TSBWR dual has both passive safety systems and active safety systems. They are also shared between the two reactors. A lot of sharing between the dual reactors enables significant cost savings accompanied by the power increase up to 3,400 MWe. Although the TSBWR dual consists of two reactors, the simplified cylindrical configuration of the key structures and reduction of the R/B height can minimize the plant construction period. The TSBWR dual provides a concept with which we can challenge to construct a dual BWR plant in the near future. (author)

Prediction of hydrogen distribution in the reactor building in CANDU6 plant

International Nuclear Information System (INIS)

Jin, Y.; Song, Y.

2008-01-01

The CANDU plants have a lot of zircaloy. The fuel cladding, calandria tubes and pressure tubes are made of zircaloy. The zircaloy can be oxidized and hydrogen is generated during severe accident progression. The detonation or deflagration to detonation transition (DDT) due to hydrogen combustion may occur if the local hydrogen concentration or global hydrogen concentration exceeds certain value. The detonation may result in the rupture of the reactor building. The inside of the reactor building of CANDU plants is complex. So prediction of hydrogen distribution in the reactor building is important. This prediction is made using ISAAC code and GOTHIC code. ISAAC code partitioned the reactor building in to 7 compartments. GOTHIC code modeled the CANDU6 reactor building using 12 nodes. The hydrogen concentrations in the various compartments in the reactor building are compared. GOTHIC code slightly underpredicts hydrogen concentration in the F/M rooms than ISAAC code, but trend is same. The hydrogen concentration in the boiler room and the moderator room shows almost same as for both codes. (author)

Clinch River Breeder Reactor Plant: a building block in nuclear technology

International Nuclear Information System (INIS)

McCormack, M.

1979-01-01

Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP

Clinch River Breeder Reactor Plant: a building block in nuclear technology

Energy Technology Data Exchange (ETDEWEB)

McCormack, M.

1979-01-01

Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP.

The working lifetime of nuclear power plants and new types of power reactors

International Nuclear Information System (INIS)

Bataille, Ch.; Birraux, C.

2003-01-01

The report on the working lifetime of nuclear power plants and new reactor types, by Mr Christian Bataille, deputy for the Nord, and Mr Claude Birraux, deputy for Haute-Savoie as well as President of the Office, supplements the studies carried out by the Parliamentary Office on the Safety of Nuclear Installations and Radioactive Wastes: it examines the remaining working life of the EDF nuclear power plants and the current status of projects that might, if circumstances were right, replace the reactors at present in service. The report investigates the different physical and other factors that influence the ageing of nuclear power plants and tackles the question of whether the design life of 40 years could be exceeded in practice. The whole issue of French nuclear power plant is put in perspective and compared with the situation of nuclear plants in Finland, Sweden, Germany and the United States, from the technical and regulatory standpoints. Believing that any attempt to optimise the working lifetime of the power plants currently in service must be accompanied by simultaneous moves aimed at their replacement, Messrs. Christian Bataille and Claude Birraux go on to review in detail the various light water reactor projects being proposed around the world for completion by 2015, as developments of existing models, in particular the EPR reactor of Framatome ANP, characterised by its competitiveness. They suggest that a first such reactor should be built as quickly as possible. Describing the other nuclear systems being investigated by research organisations not only in France but also in the United States and Sweden, Mrs. Christian Bataille and Claude Birraux review the objectives of these and the circumstances in which they might be developed, which would be unlikely to be before 2035 in view of the technological problems to be overcome and the industrial demonstration plants that would be needed

Fuel handling system of nuclear reactor plants

International Nuclear Information System (INIS)

Faulstich, D.L.

1991-01-01

This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

Proposed chemical plant initiated accident scenarios in a sulphur-iodine cycle plant coupled to a pebble bed modular reactor

International Nuclear Information System (INIS)

Brown, N.R.; Revankar, S.T.; Seker, V.; Downar, Th.J.

2010-01-01

In the sulphur-iodine (S-I) cycle nuclear hydrogen generation scheme the chemical plant acts as the heat sink for the very high temperature nuclear reactor (VHTR). Thus, any accident which occurs in the chemical plant must feedback to the nuclear reactor. There are many different types of accidents which can occur in a chemical plant. These accidents include intra-reactor piping failure, inter-reactor piping failure, reaction chamber failure and heat exchanger failure. Since the chemical plant acts as the heat sink for the nuclear reactor, any of these accidents induce a loss-of-heat-sink accident in the nuclear reactor. In this paper, several chemical plant initiated accident scenarios are presented. The following accident scenarios are proposed: i) failure of the Bunsen chemical reactor; ii) product flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iii) reactant flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iv) rupture of a reaction chamber. Qualitative analysis of these accident scenarios indicates that each result in either partial or total loss of heat sink accidents for the nuclear reactor. These scenarios are reduced to two types: i) discharge rate limited accidents; ii) discontinuous reaction chamber accidents. A discharge rate limited rupture of the SO 3 decomposition section of the SI cycle is proposed and modelled. Since SO 3 decomposition occurs in the gaseous phase, critical flow out of the rupture is calculated assuming ideal gas behaviour. The accident scenario is modelled using a fully transient control volume model of the S-I cycle coupled to a THERMIX model of a 268 MW pebble bed modular reactor (PBMR-268) and a point kinetics model. The Bird, Stewart and Lightfoot source model for choked gas flows from a pressurised chamber was utilised as a discharge rate model. A discharge coefficient of 0.62 was assumed. Feedback due to the rupture is observed in the nuclear

Plant-scale anodic dissolution of unirradiated N-Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1995-01-01

Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electro-refiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electro-refiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electro-refiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour. (author)

Reactor power control device in BWR power plant

International Nuclear Information System (INIS)

Kurosawa, Tsuneo.

1997-01-01

The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

The challenge of introducing high-temperature reactor plants onto the international power plant market

International Nuclear Information System (INIS)

Bogen, J.; Stoelzl, D.

1987-01-01

Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

The challenge of introducing high-temperature-reactor plants onto the international power plant market

International Nuclear Information System (INIS)

Bogen, J.; Stoelzl, D.

1988-01-01

Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

Operational experience with propulsion nuclear plants

International Nuclear Information System (INIS)

Polunichev, V.

2000-01-01

Russia possesses a powerful icebreaker transport fleet which offers a solution for important socio-economic tasks of the country's northern regions by maintaining a year-round navigation along the Arctic Sea route. The total operating record of the propulsion nuclear reactors till now exceeds 150 reactor-years, their main equipment items operating life amounted to 120,000 h. Progressive design-constructional solutions being perfected continuously during 40 years of nuclear-powered ships creation in Russia and well proven technology of all components used in the marine nuclear reactors give grounds to recommend marine Nuclear Steam Supply Systems (NSSSs) of KLT-40 type as energy sources for heat and power cogeneration plants and sea water desalination complexes, particularly as floating installations. Co-generation stations are considered for deployment in the extreme north of Russia. Nuclear floating desalination complexes can be used for drinkable water production in coastal regions of Northern Africa, the Near East, India etc. (author)

High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

Science.gov (United States)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

Capital cost: pressurized water reactor plant. Commerical electric power cost studies

International Nuclear Information System (INIS)

1977-06-01

The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate

Systematic simulation of a tubular recycle reactor on the basis of pilot plant experiments

Energy Technology Data Exchange (ETDEWEB)

Paar, H; Narodoslawsky, M; Moser, A [Technische Univ., Graz (Austria). Inst. fuer Biotechnologie, Mikrobiologie und Abfalltechnologie

1990-10-10

Systematic simulatiom may decisively help in development and optimization of bioprocesses. By applying simulation techniques, optimal use can be made of experimental data, decreasing development costs and increasing the accuracy in predicting the behavior of an industrial scale plant. The procedure of the dialogue between simulation and experimental efforts will be exemplified in a case study. Alcoholic fermentation of glucose by zymomonas mobilis bacteria in a gasified turbular recycle reactor was studied first by systematic simulation, using a computer model based solely on literature data. On the base of the results of this simulation, a 0.013 m{sup 3} pilot plant reactor was constructed. The pilot plant experiments, too, were based on the results of the systematic simulation. Simulated and experimental data were well in agreement. The pilot plant experiments reiterated the trends and limits of the process as shown by the simulation results. Data from the pilot plant runs were then used to improve the simulation model. This improved model was subsequently used to simulate the performances of an industrial scale plant. The results of this simulation are presented. They show that the alcohol fermentation in a tubular recycle reactor is potentially advantageous to other reactor configurations, especially to continuous stirred tanks. (orig.).

Uranium-236 in light water reactor spent fuel recycled to an enriching plant

International Nuclear Information System (INIS)

de la Garza, A.

1977-01-01

The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant

Basic safety principles of KLT-40C reactor plants

International Nuclear Information System (INIS)

Beliaev, V.; Polunichev, V.

2000-01-01

The KLT-40 NSSS has been developed for a floating power block of a nuclear heat and power station on the basis of ice-breaker-type NSSS (Nuclear Steam Supply System) with application of shipbuilding technologies. Basic reactor plant components are pressurised water reactor, once-through coil-type steam generator, primary coolant pump, emergency protection rod drive mechanisms of compensate group-electromechanical type. Basic RP components are incorporated in a compact steam generating block which is arranged within metal-water shielding tank's caissons. Domestic regulatory documents on safety were used for the NSSS design. IAEA recommendations were also taken into account. Implementation of basic safety principles adopted presently for nuclear power allowed application of the KLT-40C plant for a floating power unit of a nuclear co-generation station. (author)

Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

International Nuclear Information System (INIS)

Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

2001-01-01

Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

ELMO Bumpy Torus Reactor and power plant: conceptual design study

International Nuclear Information System (INIS)

Bathke, C.G.; Dudziak, D.J.; Krakowski, R.A.

1981-08-01

A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is presented. An emphasis is placed on those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are more generic to magnetic fusion being adapted from past, more extensive tokamak reactor designs. Similar to the latter tokamak studies, this conceptual EBTR design also emphasizes the use of conventional or near state-of-the-art engineering technology and materials. An emphasis is also placed on system accessibility, reliability, and maintainability, as these crucial and desirable characteristics relate to the unique high-aspect-ratio configuration of EBTs. Equal and strong emphasis is given to physics, engineering/technology, and costing/economics components of this design effort. Parametric optimizations and sensitivity studies, using cost-of-electricity as an object function, are reported. Based on these results, the direction for future improvement on an already attractive reactor design is identified

Aging management strategy for reactor internals of Korean nuclear power plants

International Nuclear Information System (INIS)

Hwang, Seong Sik; Kim, Soung Woo; Lee, Sam Lai; Hong, Seung Mo; Kim, Hong Pyo; Kim, Dong Jin; Lim, Yun Soo; Kim, Joung Soo; Jung, Man Kyo; Park, Jang Yul

2010-01-01

This report describes various factors on the IASCC of reactor internals in terms of fluence, stress, water chemistries and materials. Materials of each components of Korean nuclear power plants have been surveyed. A technical report for a management of reactor internals issued by EPRI was reviewed for a selection of most susceptible area among many components. Baffle former bolts are considered to be the most susceptible parts due to high irradiation level(fluence) and high tensile stress. Neutron fluence of Kori-1 and Kori-2 was calculated based on fuel exchange history, fuel performance and plant operation history. This report will be used for more advanced inspection and maintenance guidelines development by supplying inspection intervals and components (most susceptible regions) for the long term operation plants

Multifrequency tests in the EBR-II reactor plant

International Nuclear Information System (INIS)

Feldman, E.E.; Mohr, D.; Gross, K.C.

1989-01-01

A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

Science.gov (United States)

Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

2016-01-01

Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

The use of ferritic materials in light water reactor power plants

International Nuclear Information System (INIS)

Marston, T.V.

1984-01-01

This paper reviews the use of ferritic materials in LWR power plant components. The two principal types of LWR systems, the boiling water reactor (BWR) and the pressurized water reactor (PWR) are described. The evolution of the construction materials, including plates and forgings, is presented. The fabrication process for both reactors constructed with plates and forgings are described in detail. Typical mechanical properties of the reactor vessel materials are presented. Finally, one critical issue radiation embrittlement dealing with ferritic materials is discussed. This has been one of the major issues regarding the use of ferritic material in the construction of LWR pressure vessels

Plant dynamics analyses of fast reactor concept: RAPID-A without any control rod

International Nuclear Information System (INIS)

Kambe, Mitsuru

1996-01-01

Plant dynamics analyses of a fast reactor concept RAPID-A without any control rod have been demonstrated in case of reactor startup and sudden change of the primary flow rate. RAIP-A concept involves Lithium Expansion Module (LEM) for inherent reactivity feedback, Lithium Injection Module (LIM) for inherent ultimate shutdown and Lithium Release Module (LRM) for automated reactor startup. LEM consists of Quick-LEM and Slow-LEM. Slow-LEM provides with moderate reactivity addition as decreasing temperature. Quick-LEM assures quick negative reactivity feedback as increasing temperature. Plant dynamics analyses revealed that reactor power is nearly proportional to the primary flow rate even if the flow rate increases suddenly. Fully automated reactor startup from the subcritical condition has been attempted by inserting reactivity at a constant rate by LRM. Allowable rate of reactivity addition has been obtained in respect to Quick-LEM reactivity worth. (author)

Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

Energy Technology Data Exchange (ETDEWEB)

Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

2006-12-11

This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

Cost analysis of light water reactor power plants

International Nuclear Information System (INIS)

Mooz, W.E.

1978-06-01

A statistical analysis is presented of the capital costs of light water reactor (LWR) electrical power plants. The objective is twofold: to determine what factors are statistically related to capital costs and to produce a methodology for estimating these costs. The analysis in the study is based on the time and cost data that are available on U.S. nuclear power plants. Out of a total of about 60 operating plants, useful capital-cost data were available on only 39 plants. In addition, construction-time data were available on about 65 plants, and data on completed construction permit applications were available for about 132 plants. The cost data were first systematically adjusted to constant dollars. Then multivariate regression analyses were performed by using independent variables consisting of various physical and locational characteristics of the plants. The dependent variables analyzed were the time required to obtain a construction permit, the construction time, and the capital cost

Materialistic Aspects of Raising Resource of Pressurized Water Reactors for Low-Power Nuclear Plants

International Nuclear Information System (INIS)

Parshin, A.M.; Muratov, O.E.

2005-01-01

The opportunity of using ships reactors for low-power nuclear plants is considered. Some aspects of working constructional materials on cases of water-water reactors of ships nuclear units are considered. Advantages of raising resource of ships reactors are shown

The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

International Nuclear Information System (INIS)

Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

1987-01-01

The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

An overview of reactor vessel internals segmentation for nuclear plant decommissioning

International Nuclear Information System (INIS)

Litka, T.J.

1994-01-01

Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

Plant maintenance and advanced reactors issue, 2008

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal [ed.

2009-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

Energy Technology Data Exchange (ETDEWEB)

Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

Studies on environment safety and application of advanced reactor for inland nuclear power plants

International Nuclear Information System (INIS)

Wei, L.; Jie, L.

2014-01-01

To study environment safety assessment of inland nuclear power plants (NPPs), the impact of environment safety under the normal operation was researched and the environment risk of serious accidents was analyzed. Moreover, the requirements and relevant provisions of site selection between international nuclear power plant and China's are comparatively studied. The conclusion was that the environment safety assessment of inland and coastal nuclear power plant have no essential difference; the advanced reactor can meet with high criteria of environment safety of inland nuclear power plants. In this way, China is safe and feasible to develop inland nuclear power plant. China's inland nuclear power plants will be as big market for advanced reactor. (author)

Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

International Nuclear Information System (INIS)

Lear, G.; Erickson, P.B.

1975-01-01

Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

Inventory of nuclear power plants and research reactors temporary or definitively stopped in industrialized countries

International Nuclear Information System (INIS)

Clauzon, J.; Vaubert, B.

1984-12-01

This paper presents data and information on the end of the life of nuclear reactors. One deals more particularly with installations of industrialized countries. This report gives the motivations which have involved the definitive shut down of nuclear power plants and of research reactors in the concerned countries. A schedule of definitive reactor shutdowns is presented. Then, one deals with nuclear power plants of which the construction has been stopped. The reasons of these situations are also given. The temporary difficulties met during the construction or the starting of nuclear power plants these last years are mentioned. Most times, there are economical or political considerations, or safety reasons. Finally, the nuclear power plants stopped for more than two years are mentioned [fr

Natural circulation of integrated-type marine reactor at inclined attitude

International Nuclear Information System (INIS)

Iyori, Isao; Aya, Izuo; Murata, Hiroyuki; Kobayashi, Michiyuki; Nariai, Hideki

1987-01-01

A steady-state single-phase natural circulation test was performed to clarify the effect of inclination by using a model of an integrated-type marine reactor. It was found that several types of flow pattern occur in the natural circulation loop corresponding to the range of inclination angle. Stable flow rates are sustained up to near 90 0 because of the occurrence of a driving force arising from those sections of the facility which were horizontal before the inclination. It was found that the temperature distribution in the steam generator at inclined attitude depends essentially only on the elevation z. The applicability of a one-dimensional analytical model was examined. It was clarified that employment of detailed U-turn flow paths, their correlation, and temperature-distribution function of core is essential for improvement. (orig.)

Medium-Power Lead-Alloy Fast Reactor Balance-of-Plant Options

International Nuclear Information System (INIS)

Dostal, Vaclav; Hejzlar, Pavel; Todreas, Neil E.; Buongiorno, Jacopo

2004-01-01

Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature (∼550 deg. C) compared to that of light water reactors (∼300 deg. C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO 2 (S-CO 2 ) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545 deg. C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312 deg. C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO 2 cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be the S-CO 2

Plant to reduce reactor scrams and ASSET related analysis

Energy Technology Data Exchange (ETDEWEB)

Piirto, A [Teollisuuden Voima Oy (Finland)

1997-10-01

The report of events at the plant follows established rules. Basically, three categories of reports exist: first, the reactor scram report, secondly, the operational disturbance report and thirdly, the special report. The last named category covers events defined by authorities. It concentrates on safety related events, for example on failures to follow the requirements stipulated in the plant`s Technical Specifications. In general, special events are nuclear safety on the plant, the safety of the plant personnel or overall, the radiation safety in the plant`s vicinity. In the TVO`s experience feedback activity the greatest emphasis is put on events at the TVO plant. The events on the same type of plants come second. Due to limited resources, somewhat less attention is paid to events on the other types of plants. However, the experience feedback should become wider in practice so that it would be a part of everyday life in nuclear power plant operation.

Project of SVBR-75/100 reactor plant with improved safety for nuclear sources of small and medium power

International Nuclear Information System (INIS)

Dragunov, Yu. G.; Stepanov, V. S.; Klimov, N. N.; Dedul, A. V.; Bolvanchikov, S. N.; Zrodnikov, A. V.; Tolhinsky, G. I.; Komlev, O. G.

2004-01-01

As a result of the joint work performed recently by FSUE OKB Gidropress, SNC RF-IPPE and other organizations the technical feasibility is shown for creation and usage in nuclear power engineering of the unified reactor plant (RP) SVBR-75/100 with fast neutron reactor core and lead-bismuth coolant (LBC) in the primary circuit. Technical design of SVBR-75/100 reactor plant is based on the following: 50-year operation experience in development and operation of RP with LBC for nuclear submarines; experience in development and operation of fast reactor with sodium coolant; experience in optimization of LBC technology at nuclear submarines and ground-based test benches; conceptual design of SVBR-75 reactor plant (for renovation of Units 2, 3 and 4 of Novovoronezh NPP). Technical solutions laid down in the basis of SVBR-75/100 reactor plant design are oriented towards the industrial basis, structural materials existing in Russia, as well as the unique LBC technology with experimental and practical support. The concept of SVBR-75/100 reactor plant safety assurance is based on the following provisions: maximum usage of inherent safety supported by physical features of fast neutron reactor, chemically inert LBC in the primary circuit, integral layout and special design solutions; maximum possible combination of normal operation and safety functions in RP systems. Small power of SVBR-75/100 RP makes it possible to manufacture the complete set of RP main equipment at the factory and delivery it to NPP site as-finished practically using any transport including railway. Possible fields of application of SVBR-75/100 reactor plants: modular NPPs of different power; renovation of NPPs with light water reactors exhausted their service life; independent nuclear power sources for different applications (ground-based nuclear water-desalinating plants, etc. )(author)

Russian spent marine fuel as a global security risk

International Nuclear Information System (INIS)

Gussgard, K.; Reistad, O.

2001-01-01

Russian marine fuel is a trans-national security concern. This paper focuses on specific technical properties of Russian marine nuclear fuel especially relevant for evaluating different aspects on nuclear proliferation, in addition to risks associated with regional environmental degradation and illegal diversion of radiological substances. Russian fresh fuel for marine reactors has been involved in several significant cases of illicit trafficking of special nuclear materials. The amount and quality of nuclear materials in Russian spent marine fuel give also reason for concern. Not less than 200 marine reactor cores are ready for having their spent fuel unloaded and preliminary stored on shore in the Far East and North West of Russia, and large amounts of spent naval fuel have been stored at Russian military bases for decades. In order to assess the security risks associated with Russian spent marine fuel, this paper discusses the material attractiveness of spent fuel from all types of Russian marine reactors. The calculations are based on a model of a light water moderated Russian icebreaker reactor. The computer tool HELIOS, used for modelling the reactor and the reactor operations, has been extensively qualified by comparisons with experimental data and international benchmark problems for reactor physics codes as well as through feedback from applications. Some of these benchmarks and studies include fuel enrichments up to 90% in Russian marine reactors. Several fuel data cases are discussed in the paper, focusing especially on: 1) early fuel designs with low initial enrichment; 2) more modern fuel designs used in third and fourth generation of Russian submarines probably with intermediate enriched fuel; and 3) marine fuel with initial enrichment levels close to weapons-grade material. In each case the fuel has been burned until k eff has reached below 1. Case 1) has been evaluated, the calculations made as basis for this paper have concentrated on fuel with

Quality assurance in technology development for The Clinch River Breeder Reactor Plant Project

International Nuclear Information System (INIS)

Anderson, J.W.

1980-01-01

The Clinch River Breeder Reactor Plant Project is the nation's first large-scale demonstration of the Liquid Metal Fast Breeder Reactor (LMFBR) concept. The Project has established an overall program of plans and actions to assure that the plant will perform as required. The program has been established and is being implemented in accordance with Department of Energy Standard RDT F 2-2. It is being applied to all parts of the plant, including the development of technology supporting its design and licensing activity. A discussion of the program as it is applied to development is presented

Operational margin monitoring system for boiling water reactor power plants

International Nuclear Information System (INIS)

Fukutomi, S.; Takigawa, Y.

1992-01-01

This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

Safety Philosophy in Process Heat Plants Coupled to High Temperature Reactors

International Nuclear Information System (INIS)

Brown, Nicholas R.; Revankar, Shripad T.

2011-01-01

With the future availability of fossil fuel resources in doubt, high temperature nuclear reactors have the potential to be an important technology in the near term. Due to a high coolant outlet temperature, high temperature reactors (HTR) can be used to drive chemical plants that directly utilize process heat. Additionally, the high temperature improves the thermodynamic efficiency of the energy utilization. Many applications of high temperature reactors exist as a thermal driving vector for endothermic chemical process plants. Hydrogen generation using the General Atomics (GA) sulfur iodine (SI) cycle is one promising application of high temperature nuclear heat. The main chemical reactions in the SI cycle are: 1. I 2 +SO 2 + 2H 2 O → 2HI + H 2 SO 4 (Bunsen reaction) 2. H 2 SO 4 → H 2 O + SO 2 + 1/2O 2 (Sulfuric acid decomposition) 3. 2HI → H 2 + I 2 (Hydrogen Iodide decomposition). With the exception of hydrogen and oxygen, all relevant reactants are recycled within the process. However, there are many unresolved safety and operational issues related to implementation of such a coupled plant

Fuel handling grapple for nuclear reactor plants

International Nuclear Information System (INIS)

Rousar, D.L.

1992-01-01

This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

Use of reactor plants of enhanced safety for sea water desalination, industrial and district heating

International Nuclear Information System (INIS)

Panov, Yu.; Polunichev, V.; Zverev, K.

1997-01-01

Russian designers have developed and can deliver nuclear complexes to provide sea water desalination, industrial and district heating. This paper provides an overview of these designs utilizing the ABV, KLT-40 and ATETS-80 reactor plants of enhanced safety. The most advanced nuclear powered water desalination project is the APVS-80. This design consists of a special ship equipped with the distillation desalination plant powered at a level of 160 MW(th) utilizing the type KLT-40 reactor plant. More than 20 years of experience with water desalination and reactor plants has been achieved in Aktau and Russian nuclear ships without radioactive contamination of desalinated water. Design is also proceeding on a two structure complex consisting of a floating nuclear power station and a reverse osmosis desalination plant. This new technology for sea water desalination provides the opportunity to considerably reduce the specific consumption of power for the desalination of sea water. The ABV reactor is utilized in the ''Volnolom'' type floating nuclear power stations. This design also features a desalinator ship which provides sea water desalination by the reverse osmosis process. The ATETS-80 is a nuclear two-reactor cogeneration complex which incorporates the integral vessel-type PWR which can be used in the production of electricity, steam, hot and desalinated water. (author). 9 figs

Safety problems of nuclear power plants with channel-type graphite boiling water reactors

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Vasilevskij, V.P.; Volkov, V.P.; Gavrilov, P.A.; Kramerov, A.Ya.; Kuznetsov, S.P.; Kunegin, E.P.; Rybakov, N.Z.

1977-01-01

Construction of nuclear power plants in a highly populated region near large industrial centres necessitates to pay a special attention to their nuclear and radiation safety. Safety problems of nuclear reactor operation are discussed, in particular, they are: reliable stoppage of fission chain reaction at any emergency cases; reliable core cooling with failure of various equipment; emergency core cooling with breached pipes of a circulating circuit; and prevention of radioactive coolant release outside the nuclear power plant in amount exceeding the values adopted. Channel-type water boiling reactors incorporate specific features requiring a new approach to safety operation of a reactor and a nuclear power plant. These include primarily a rather large steam volume in the coolant circuit, large amount of accumulated heat, void reactivity coefficient. Channel-type reactors characterized by fair neutron balance and flexible fuel cycle, have a series of advantages alleviating the problem of ensuring their safety. The possibility of reliable control over the state of each channel allows to replace failed fuel elements by the new ones, when operating on-load, to increase the number of circulating loops and reduce the diameter of main pipelines, simplifies significantly the problem of channel emergency cooling and localization of a radioactive coolant release from a breached circuit. The concept of channel-type reactors is based on the solution of three main problems. First, plant safety should be assured in emergency switch off of separate units and, if possible, energy conditions should be maintained, this is of particular importance considering the increase in unit power. Second, the system of safety and emergency cooling should eliminate a great many failures of fuel elements in case of potential breaches of any tube in the circulating circuit. Finally, rugged boxes and localizing devices should be provided to exclude damage of structural elements of the nuclear power

Antifungal potential of marine sponge extract against plant and fish pathogenic fungi

Digital Repository Service at National Institute of Oceanography (India)

PrabhaDevi; Ravichandran, S.; Ribeiro, M.; Ciavatta, M.L.

scope for rediscovering compounds with antimicrobial activity. This study screens extracts (Diethyl Ether and Butanol) of a marine red-Sea sponge Negombata magnifica for invitro fungicidal activity against 10 plant and 3 fish pathogens. Fungicidal...

Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

Power-up of Fugen reactor and development of demonstration plant

International Nuclear Information System (INIS)

Sawai, Sadamu; Akebi, Michio; Yazaki, Akira.

1979-06-01

The Fugen Nuclear Power Station is the 165 MWe prototype plant characterized by heavy water-moderated, boiling light water-cooled, pressure tube type, and was developed by the Power Reactor and Nuclear Fuel Development Corporation, Japan. The plant went into commercial operation on March 20, 1979, in Tsuruga, Fukui Prefecture. Some delay in the overall schedule occurred due to the shortage of cement caused by the oil crisis, more stringent regulations as the result of stress corrosion cracking experienced in BWRs, and design modifications. All functional tests, the final check-up of the whole plant, and remaining modifying works had been completed by March 10, 1978. The minimum criticality was achieved with 22 mixed oxide fuel assemblies on March 20, 1978. Thereafter, the tests on reactor physics, plant dynamics, the performances of components and systems, and radiation and water chemistry have been carried out. 5 MWe was sent to grid system for the first time on July 29, 1978. The commercial operation of the plant was licenced by the Government on March 30, 1979. The conceptual design of the 600 MWe demonstration plant was finished in early 1979, and the detailed design is to be carried out in 1979 and 1980. The main design principle was incorporated in the conceptual design, but some modifications are to be made to reduce the power cost and to facilitate the easy maintenance. (Kako, I.)

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-05-01

The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

Lessons learned from the licensing process for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Dickson, P.W.; Clare, G.H.

1991-01-01

This paper presents the experience of licensing a specific liquid-metal fast breeder reactor (LMFBR), the Clinch River Breader Reactor Plant (CRBRP). It was a success story in that the licensing process was accomplished in a very short time span. The actions of the applicant and the actions of the US Nuclear Regulatory Commission (NRC) in response are presented and discussed to provide guidance to future efforts to license unconventional reactors. The history is told from the perspective of the authors. As such, some of the reasons given for success or lack of success are subjective interpretations. Nevertheless, the authors' positions provided them an excellent viewpoint to make these judgements. During the second phase of the licensing process, they were the CRBRP Technical Director and the Licensing Manager, respectively, for the Westinghouse Electric Corporation, the prime contractor for the reactor plant

Carbon-14 in reactor plant water

International Nuclear Information System (INIS)

Knowles, G.K.

1979-01-01

The method for the analysis of 14 C in reactor plant water and various waste streams previously used at the Idaho National Engineering Laboratory has been shown to be ineffective for samples which contain organic compounds. The previous method consisted of acidification and refluxing of the sample, precipitation of the liberated CO 2 , and subsequent analysis by the liquid scintillation method. The method was simple but it did not convert all compounds containing 14 C in the sample to CO 2 . The new method, while it is based on the previous method, has been improved by employing a strong oxidant, potassium persulfate and silver nitrate, for more complete oxidation of the organics to CO 2 . The new method yields 14 C values that have typically been one to two orders of magnitude higher than the values obtained using the former method. This indicates that most of the 14 C present in the current reactor water samples being analyzed is associated with trace amounts of organics

Balance of plant design issues for small reactors in Canada

International Nuclear Information System (INIS)

Harvel, G.; Meneley, D.

2014-01-01

Internationally, several companies are exploring design and development of Small Modular Reactors (SMR) ranging in power from 10 MWe to 300 MWe. While the designs are proceeding, the main issue at hand is finding a site for deployment of the first unit. Connection to existing well established grids is currently not competitive in part due to First of a Kind (FOAK) costs. As such, many vendors are exploring unique and remote applications where FOAK costs are not as significant a concern. One of the major assumptions in the design process usually followed is that the major effort needs to concentrate on reactor core development. While the reactor core is important, costs associated with the balance of plant and operations of the unit are likely to play an important role in the final decision of purchase. In this work, a series of conceptual designs is performed for the support systems of a small modular reactor by successive teams of undergraduate students working over semester long periods during a 3 year period. The goal of this process is to determine to what extent current technology for the balance of plant supports the development of a cost effective SMR. Each system is given to a team with an open set of criteria for design. At the completion of the design exercise, an open discussion with the teams is held regarding the staffing requirements for an SMR. The results are preliminary and reflect the open nature of the exercise. That said, the results indicate that for an SMR to be truly competitive, significant innovation is required in addressing the supporting systems of the plant. (author)

Balance of plant design issues for small reactors in Canada

Energy Technology Data Exchange (ETDEWEB)

Harvel, G.; Meneley, D., E-mail: Glenn.Harvel@uoit.ca, E-mail: dan.meneley@sympatico.ca [Univ. of Ontario Inst. of Tech.y, Oshawa, ON (Canada)

2014-07-01

Internationally, several companies are exploring design and development of Small Modular Reactors (SMR) ranging in power from 10 MWe to 300 MWe. While the designs are proceeding, the main issue at hand is finding a site for deployment of the first unit. Connection to existing well established grids is currently not competitive in part due to First of a Kind (FOAK) costs. As such, many vendors are exploring unique and remote applications where FOAK costs are not as significant a concern. One of the major assumptions in the design process usually followed is that the major effort needs to concentrate on reactor core development. While the reactor core is important, costs associated with the balance of plant and operations of the unit are likely to play an important role in the final decision of purchase. In this work, a series of conceptual designs is performed for the support systems of a small modular reactor by successive teams of undergraduate students working over semester long periods during a 3 year period. The goal of this process is to determine to what extent current technology for the balance of plant supports the development of a cost effective SMR. Each system is given to a team with an open set of criteria for design. At the completion of the design exercise, an open discussion with the teams is held regarding the staffing requirements for an SMR. The results are preliminary and reflect the open nature of the exercise. That said, the results indicate that for an SMR to be truly competitive, significant innovation is required in addressing the supporting systems of the plant. (author)

Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

International Nuclear Information System (INIS)

Yamada, Fumiaki; Mori, Takero

2005-01-01

In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

Floating nuclear heat. And power station 'Pevec' with KLT-40S type reactor plant for remote regions of Russia

International Nuclear Information System (INIS)

Veshnyakov, K.B.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

2000-01-01

Floating small nuclear power plants power for local energy systems of littoral regions of Russia, located far from central energy system, open a new line in nuclear power development. Designing a floating power unit of a lead nuclear heat and power generating station for port Pevec at the Chuckchee national district is currently nearing completion. Most labor-intensive components are being manufactured. The co-generation NPP Pevec is to be created on the basis of a floating power unit with KLT-40S type reactor plant. KLT-40S reactor plant is based on similar propulsion plants, verified at operation of Russia's nuclear-powered civil ships, evolutionary improved by elimination of 'weak points' revealed during its prototypes operation or on the basis of safety analysis. KLT-40S reactor plant uses the most wide-spread and developed in the world practice PWR-type reactor. KLT-40S meets contemporary national and international requirements imposed to future reactor plants. The NHPS description, its main technical-economic data, environmental safety indices, basic characteristics of KLT-40S reactor plant are presented. Prospects of small NPPs utilization outside Russia, particularly as an energy source for sea water desalination, are considered. (author)

Plant to reduce reactor scrams and ASSET related analysis

International Nuclear Information System (INIS)

Piirto, A.

1997-01-01

The report of events at the plant follows established rules. Basically, three categories of reports exist: first, the reactor scram report, secondly, the operational disturbance report and thirdly, the special report. The last named category covers events defined by authorities. It concentrates on safety related events, for example on failures to follow the requirements stipulated in the plant's Technical Specifications. In general, special events are nuclear safety on the plant, the safety of the plant personnel or overall, the radiation safety in the plant's vicinity. In the TVO's experience feedback activity the greatest emphasis is put on events at the TVO plant. The events on the same type of plants come second. Due to limited resources, somewhat less attention is paid to events on the other types of plants. However, the experience feedback should become wider in practice so that it would be a part of everyday life in nuclear power plant operation

Development of a built-in type Control Rod Drive Mechanism (CRDM) for Advanced Marine Reactor X (MRX)

International Nuclear Information System (INIS)

Ishizaka, Y.; Iida, H.; Yamaji, A.

1992-01-01

For realization of the next generation Advanced Marine Reactor X(MRX) with higher safety, design studies and basic experiments have been done on the built-in type Control Rod Drive Mechanism (CRDM). The concept has been made clear of the CRDM that can be placed inside the reactor vessel and fits best to the MRX - an integrated-type PWR. In particular, the design has almost been completed for the driving motor and the latch magnet, which are the core of this CRDM. It is expected that the required performance can be assured even if there are losses due to the high temperature effect. (author)

Topics to be covered in safety analysis reports for nuclear power plants with pressurized water reactors or boiling water reactors in the F.R.G

International Nuclear Information System (INIS)

Kohler, H.A.G.

1977-01-01

This manual aims at defining the standards to be used in Safety Analysis Reports for Nuclear Power Plants with Pressurized Water Reactors or Boiling Water Reactors in the Federal Republic of Germany. The topics to be covered are: Information about the site (geographic situation, settlement, industrial and military facilities, transport and communications, meteorological conditions, geological, hydrological and seismic conditions, radiological background), description of the power plant (building structures, safety vessel, reactor core, cooling system, ventilation systems, steam power plant, electrical facilities, systems for measurement and control), indication of operation (commissioning, operation, safety measures, radiation monitoring, organization), incident analysis (reactivity incidents, loss-of-coolant incidents, external impacts). (HP) [de

Annealing the reactor vessel at the Palisades Plant

International Nuclear Information System (INIS)

Fenech, R.A.

1996-01-01

In the way of background, Palisades was licensed in 1967 and went commercial in 1971. Jumping to two years ago, we faced at that time three issues that challenged our ability to operate to end-of-license, which would be 2007 without any extensions. The three items were regulatory performance, economic performance, and reactor vessel embrittlement. We had not been operating the plant with the kind of conservative decisions and with the kind of safety margins that one is expected to operate a plant in the United States at this time. Our economic performance was not satisfactory in that our capacity factor was low and our costs high. In the area of reactor vessel embrittlement, our analysis showed that we would reach the NRC screening criteria for embrittlement in the year 2004. Over the last two years, we have made significant improvements in the first two areas. Our decision-making has changed. Our performance, especially over the last year and a half, has been excellent. In addition, we have gotten our capacity factors up and our costs under control. Clearly, sustained performance is what is going to carry the day but from what we can see and from where we are, we are in more of a maintenance-of-performance than in a turn-around situation. On the other hand, in the area of reactor vessel embrittlement, about a year and a half ago we had a bit of a setback. We had taken material from retired steam generators that had welds identical to the welds in our reactor vessel. When we analyzed the welds from our steam generators, we were given some surprises about the chemistry makeup. When we applied the new information to our analysis, we changed the date on which we would reach our screening criteria from 2004 to late 1999

Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

International Nuclear Information System (INIS)

Wati, Nurokhim

2008-01-01

Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

Application of fault tree methodology to modeling of the AP1000 plant digital reactor protection system

International Nuclear Information System (INIS)

Teolis, D.S.; Zarewczynski, S.A.; Detar, H.L.

2012-01-01

The reactor trip system (RTS) and engineered safety features actuation system (ESFAS) in nuclear power plants utilizes instrumentation and control (IC) to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations. During normal operating conditions, various plant parameters are continuously monitored to assure that the plant is operating in a safe state. In response to deviations of these parameters from pre-determined set points, the protection system will initiate actions required to maintain the reactor in a safe state. These actions may include shutting down the reactor by opening the reactor trip breakers and actuation of safety equipment based on the situation. The RTS and ESFAS are represented in probabilistic risk assessments (PRAs) to reflect the impact of their contribution to core damage frequency (CDF). The reactor protection systems (RPS) in existing nuclear power plants are generally analog based and there is general consensus within the PRA community on fault tree modeling of these systems. In new plants, such as AP1000 plant, the RPS is based on digital technology. Digital systems are more complex combinations of hardware components and software. This combination of complex hardware and software can result in the presence of faults and failure modes unique to a digital RPS. The United States Nuclear Regulatory Commission (NRC) is currently performing research on the development of probabilistic models for digital systems for inclusion in PRAs; however, no consensus methodology exists at this time. Westinghouse is currently updating the AP1000 plant PRA to support initial operation of plants currently under construction in the United States. The digital RPS is modeled using fault tree methodology similar to that used for analog based systems. This paper presents high level descriptions of a typical analog based RPS and of the AP1000 plant digital RPS. Application of current fault

Multi-stage-flash desalination plants of relative small performance with integrated pressurized water reactors as a nuclear heat source

International Nuclear Information System (INIS)

Petersen, G.; Peltzer, M.

1977-01-01

In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination plants with a performance in the range of 10 000 to 80 000 m 3 distillate per day heated by a nuclear reactor are investigated. The reactor concept is similar to the Integrated Pressurized Water Reactor (IPWR) of the nuclear ship OTTO HAHN. The design study shows that IPWR systems have specific advantages up to 200 MWth compared to other reactor types at least being adapted for single- and dual-purpose desalination plants. The calculated costs of the desalinated water show that due to fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (author)

Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

International Nuclear Information System (INIS)

Kuruc, J.; Matel, L.

2007-01-01

The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

Energy Technology Data Exchange (ETDEWEB)

Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

2007-03-21

In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

Clinch river breeder reactor plant steam generator water quality

Energy Technology Data Exchange (ETDEWEB)

Van Hoesen, D; Lowe, P A

1975-07-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

Clinch river breeder reactor plant steam generator water quality

International Nuclear Information System (INIS)

Van Hoesen, D.; Lowe, P.A.

1975-01-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

Energy Technology Data Exchange (ETDEWEB)

Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

1999-05-15

The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

1980 nuclear plant survey: no reactors sold; more cancellations

International Nuclear Information System (INIS)

Friedlander, G.D.

1980-01-01

No sales were reported in 1979 by any of the big four reactor suppliers. Three cancellations were reported and construction was suspended on the Jersey Central Power and Light Co.'s Forked River unit. Since last year's survey, the commercial operation dates of about 80 units have been postponed from one year to indefinitely, and nuclear commitments are down from last year's 195 units to 193 units. Presently, there are 72 plants on line, with a capacity of more than 53,000 MW. A resumption of new reactor orders is expected in either late 1980 or early 1981

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

Energy Technology Data Exchange (ETDEWEB)

1993-05-13

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

Ocean acidification and the loss of phenolic substances in marine plants.

Directory of Open Access Journals (Sweden)

Thomas Arnold

Full Text Available Rising atmospheric CO(2 often triggers the production of plant phenolics, including many that serve as herbivore deterrents, digestion reducers, antimicrobials, or ultraviolet sunscreens. Such responses are predicted by popular models of plant defense, especially resource availability models which link carbon availability to phenolic biosynthesis. CO(2 availability is also increasing in the oceans, where anthropogenic emissions cause ocean acidification, decreasing seawater pH and shifting the carbonate system towards further CO(2 enrichment. Such conditions tend to increase seagrass productivity but may also increase rates of grazing on these marine plants. Here we show that high CO(2 / low pH conditions of OA decrease, rather than increase, concentrations of phenolic protective substances in seagrasses and eurysaline marine plants. We observed a loss of simple and polymeric phenolics in the seagrass Cymodocea nodosa near a volcanic CO(2 vent on the Island of Vulcano, Italy, where pH values decreased from 8.1 to 7.3 and pCO(2 concentrations increased ten-fold. We observed similar responses in two estuarine species, Ruppia maritima and Potamogeton perfoliatus, in in situ Free-Ocean-Carbon-Enrichment experiments conducted in tributaries of the Chesapeake Bay, USA. These responses are strikingly different than those exhibited by terrestrial plants. The loss of phenolic substances may explain the higher-than-usual rates of grazing observed near undersea CO(2 vents and suggests that ocean acidification may alter coastal carbon fluxes by affecting rates of decomposition, grazing, and disease. Our observations temper recent predictions that seagrasses would necessarily be "winners" in a high CO(2 world.

Modeling and simulation of pressurized water reactor power plant

International Nuclear Information System (INIS)

Wang, S.J.

1983-01-01

Two kinds of balance of plant (BOP) models of a pressurized water reactor (PWR) system are developed in this work - the detailed BOP model and the simple BOP model. The detailed model is used to simulate the normal operational performance of a whole BOP system. The simple model is used to combine with the NSSS model for a whole plant simulation. The trends of the steady state values of the detailed model are correct and the dynamic responses are reasonable. The simple BOP model approach starts the modelling work from the overall point of view. The response of the normalized turbine power and the feedwater inlet temperature to the steam generator of the simple model are compared with those of the detailed model. Both the steady state values and the dynamic responses are close to those of the detailed model. The simple BOP model is found adequate to represent the main performance of the BOP system. The simple balance of plant model was coupled with a NSSS model for a whole plant simulation. The NSSS model consists of the reactor core model, the steam generator model, and the coolant temperature control system. A closed loop whole plant simulation for an electric load perturbation was performed. The results are plausible. The coupling effect between the NSSS system and the BOP system was analyzed. The feedback of the BOP system has little effect on the steam generator performance, while the performance of the BOP system is strongly affected by the steam flow rate from the NSSS

Noble gas binary mixtures for gas-cooled reactor power plants

International Nuclear Information System (INIS)

El-Genk, Mohamed S.; Tournier, Jean-Michel

2008-01-01

This paper examines the effects of using noble gases and binary mixtures as reactor coolants and direct closed Brayton cycle (CBC) working fluids on the performance of terrestrial nuclear power plants and the size of the turbo-machines. While pure helium has the best transport properties and lowest pumping power requirement of all noble gases and binary mixtures, its low molecular weight increases the number of stages of the turbo-machines. The heat transfer coefficient for a He-Xe binary mixture having a molecular weight of 15 g/mole is 7% higher than that of helium, and the number of stages in the turbo-machines is 24-30% of those for He working fluid. However, for the same piping and heat exchange components design, the loop pressure losses with He-Xe are ∼3 times those with He. Consequently, for the same reactor exit temperature and pressure losses in piping and heat exchange components, the higher pressure losses in the nuclear reactor decrease the net peak efficiency of the plant with He-Xe working fluid (15 g/mole) by a little more than ∼2% points, at higher cycle compression ratio than with He working fluid

Rock siting of nuclear power plants from a reactor safety standpoint

International Nuclear Information System (INIS)

1975-11-01

The study has aimed at surveying the advantages and disadvantages of a rock sited nuclear power plant from a reactor safety standpoint. The studies performed are almost entirely concentrated on the BWR alternative. The design of a nuclear power plant in rock judged most appropriate has been studied in greater detail, and a relatively extensive safety analysis has been made. It is found that the presented technical design of the rock sited alternative is sufficiently advanced to form a basis for further projecting treatment. The chosen technical design of the reactor plant demands a cavern with a 45-50 metre span. Caverns without strengthening efforts with such spans are used in mines, but have no previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant. When locating a rock sited nuclear power plant, the same criteria must be considered as for an above ground plant, with additional stronger demands for rock quality. The presented rock sited nuclear power plant has been assessed to cost 20 % more in total construction costs than a corresponding above ground plant. The motivations for rock siting also depend on whether a condensing plant for only electricity production, or a plant for combined power production and district heating, is considered. The latter would under certain circumstances make rock siting look more attractive. (author)

RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

International Nuclear Information System (INIS)

Kawai, H.

2001-01-01

JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

Radioactive waste in the marine environment

International Nuclear Information System (INIS)

Lowman, F.G.

1975-01-01

The introduction of radionuclides into near-shore marine areas is considered in relation to Federal Regulations and to the hazards posed to man and to marine organisms living near the site of introduction of the contaminants. The use of the critical pathway concept, for evaluating hazards to man from radionuclides in his food, is discussed and the specific activity concept is applied to demonstrate the degree of hazard to man which could result from normal reactor operation using sea water for the coolant. The relative hazards to marine organisms from naturally occurring and reactor-produced radionuclides are also evaluated. (U.S.)

Membrane bio-reactor for textile wastewater treatment plant upgrading.

Science.gov (United States)

Lubello, C; Gori, R

2005-01-01

Textile industries carry out several fiber treatments using variable quantities of water, from five to forty times the fiber weight, and consequently generate large volumes of wastewater to be disposed of. Membrane Bio-reactors (MBRs) combine membrane technology with biological reactors for the treatment of wastewater: micro or ultrafiltration membranes are used for solid-liquid separation replacing the secondary settling of the traditional activated sludge system. This paper deals with the possibility of realizing a new section of one existing WWTP (activated sludge + clariflocculation + ozonation) for the treatment of treating textile wastewater to be recycled, equipped with an MBR (76 l/s as design capacity) and running in parallel with the existing one. During a 4-month experimental period, a pilot-scale MBR proved to be very effective for wastewater reclamation. On average, removal efficiency of the pilot plant (93% for COD, and over 99% for total suspended solids) was higher than the WWTP ones. Color was removed as in the WWTP. Anionic surfactants removal of pilot plant was lower than that of the WWTP (90.5 and 93.2% respectively), while the BiAS removal was higher in the pilot plant (98.2 vs. 97.1). At the end cost analysis of the proposed upgrade is reported.

Life time of nuclear power plants and new types of reactors

International Nuclear Information System (INIS)

2003-05-01

This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

EPRI's nuclear power plant instrumentation and control program and its applicability to advanced reactors

International Nuclear Information System (INIS)

Naser, J.; Torok, R.; Wilkinson, D.

1997-01-01

I ampersand C systems in nuclear power plants need to be upgraded over the lifetime of the plant in a reliable and cost-effective manner to replace obsolete equipment, to reduce O ampersand M costs, to improve plant performance, and to maintain safety. This applies to operating plants now and will apply to advanced reactors in the future. The major drivers for the replacement of the safety, control, and information systems in nuclear power plants are the obsolescence of the existing hardware and the need for more cost-effective power production. Competition between power producers is dictating more cost-effective power production. The increasing O ampersand M costs to maintain systems experiencing obsolescence problems is counter to the needs for more cost-effective power production and improved competitiveness. This need for increased productivity applies to government facilities as well as commercial plants. Increasing competition will continue to be a major factor in the operation of both operating plants and advanced reactors. It will continue to dictate the need for improved productivity and cost-effectiveness. EPRI and its member nuclear utilities are working together on an industry wide I ampersand C Program to address I ampersand C issues and to develop cost-effective solutions. A majority of the I ampersand C products and demonstrations being developed under this program will benefit advanced reactors in both the design and operational phases of their life cycle as well as it will benefit existing plants. 20 refs

Capital cost: pressurized water reactor plant. Commercial electric power cost studies

Energy Technology Data Exchange (ETDEWEB)

1977-06-01

The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

Capital cost: pressurized water reactor plant. Commercial electric power cost studies

International Nuclear Information System (INIS)

1977-06-01

The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description

A dynamic model of the reactor coolant system flow for KMRR plant simulation

International Nuclear Information System (INIS)

Rhee, B.W.; Noh, T.W.; Park, C.; Sim, B.S.; Oh, S.K.

1990-01-01

To support computer simulation studies for reactor control system design and performance evaluation, a dynamic model of the reactor coolant system (RCS) and reflector cooling system has been developed. This model is composed of the reactor coolant loop momentum equation, RCS pump dynamic equation, RCS pump characteristic equation, and the energy equation for the coolant inside the various components and piping. The model is versatile enough to simulate the normal steady-state conditions as well as most of the anticipated flow transients without pipe rupture. This model has been successfully implemented as the plant simulation code KMRRSIM for the Korea Multi-purpose Research Reactor and is now under extensive validation testing. The initial stage of validation has been comparison of its result with that of already validated, more detailed reactor system transient codes such as RELAP5. The results, as compared to the predictions by RELAP5 simulation, have been generally found to be very encouraging and the model is judged to be accurate enough to fulfill its intended purpose. However, this model will continue to be validated against other plant's data and eventually will be assessed by test data from KMRR

Nonlinear Dynamic Model of Power Plants with Single-Phase Coolant Reactors

International Nuclear Information System (INIS)

Vollmer, H.

1968-12-01

The traditional way of developing dynamic models for a specific nuclear power plant and for specific purpose seems rather uneconomical, as much of the information often can not be utilized if the plant design or the required accuracy of the calculation is desired to be changed. It is therefore suggested that the model development may be made more systematic, general and flexible by - applying the 'box of bricks' system, where the main components of a nuclear power plant are treated separately and combined afterwards according to a given flow scheme, - a dynamic determination of the components which is as general as possible without taking into account those details which have a minor influence on the overall dynamics, - providing approximations of the more rigorous solution sufficient to meet the user s requirements on accuracy, - proper use of computers. A dynamic model for single-phase coolant reactor plants is established along these lines. By separation of the nonlinear and linear parts of the system, application of Laplace transformation and proper approximations, and the use of a hybrid computer it seems possible to determine the (nonlinear) dynamic behaviour of such a plant for perturbations which are not so large that phase changes of physical parameters occur, e. g. fuel does not melt. The model is applied to a steam cooled fast reactor power plant

Nonlinear Dynamic Model of Power Plants with Single-Phase Coolant Reactors

Energy Technology Data Exchange (ETDEWEB)

Vollmer, H

1968-12-15

The traditional way of developing dynamic models for a specific nuclear power plant and for specific purpose seems rather uneconomical, as much of the information often can not be utilized if the plant design or the required accuracy of the calculation is desired to be changed. It is therefore suggested that the model development may be made more systematic, general and flexible by - applying the 'box of bricks' system, where the main components of a nuclear power plant are treated separately and combined afterwards according to a given flow scheme, - a dynamic determination of the components which is as general as possible without taking into account those details which have a minor influence on the overall dynamics, - providing approximations of the more rigorous solution sufficient to meet the user s requirements on accuracy, - proper use of computers. A dynamic model for single-phase coolant reactor plants is established along these lines. By separation of the nonlinear and linear parts of the system, application of Laplace transformation and proper approximations, and the use of a hybrid computer it seems possible to determine the (nonlinear) dynamic behaviour of such a plant for perturbations which are not so large that phase changes of physical parameters occur, e. g. fuel does not melt. The model is applied to a steam cooled fast reactor power plant.

Events Calendar: Smithsonian Marine Ecosystems Exhibit: Smithsonian Marine

Science.gov (United States)

current Smithsonian research on the plants and animals of the Indian River Lagoon and marine environments Station (SMS) at Fort Pierce Smithsonian Marine Station at Fort Pierce Website Search Box History Modeling Ecosystems Virtual Tour Facebook Instagram Twitter SMS Home › Smithsonian Marine

Nuclear plant-aging research on reactor protection systems

International Nuclear Information System (INIS)

Meyer, L.C.

1988-01-01

This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed

Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

Development project HTR-electricity-generating plant, concept design of an advanced high-temperature reactor steam cycle plant with spherical fuel elements (HTR-K)

International Nuclear Information System (INIS)

1978-07-01

The report gives a survey of the principal work which was necessary to define the design criteria, to determine the main design data, and to design the principal reactor components for a large steam cycle plant. It is the objective of the development project to establish a concept design of an edvanced steam cycle plant with a pebble bed reactor to permit a comparison with the direct-cycle-plant and to reach a decision on the concept of a future high-temperature nuclear power plant. It is tried to establish a largerly uniform basic concept of the nuclear heat-generating systems for the electricity-generating and the process heat plant. (orig.) [de

Study on plant concept for gas cooled fast reactor

International Nuclear Information System (INIS)

Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

2003-05-01

In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

Marine environment status assessment based on macrophytobenthic plants as bio-indicators of heavy metals pollution

International Nuclear Information System (INIS)

Zalewska, Tamara; Danowska, Beata

2017-01-01

The main aim of study was to develop the environmental quality standards (EQS MP ) for selected heavy metals: Pb, Cd, Hg and Ni bioaccumulated in the tissues of marine macrophytobenthic plants: Chara baltica, Cladophora spp., Coccotylus truncatus, Furcellaria lumbricalis, Polysiphonia fucoides, Stuckenia pectinata and Zanichellia palustris, collected in designated areas of the southern Baltic Sea in period 2008–2015. The calculated concentration ratios (CR), which attained very high values: 10 4 L kg −1 for lead, 10 3 L kg −1 for nickel and mercury and even 10 5 L kg −1 for cadmium formed the basis for the determination of EQS MP values. The EQS MP values were: 26 mg kg −1 d.w. for Pb, 33 mg kg −1 d.w. for Cd, 32 mg kg −1 d.w. for Ni and 0.4 mg kg −1 d.w. for Hg. The application of macrophytobenthic plants as bioindicators in marine environment status assessment of certain areas of the Baltic Sea is also described in the paper. - Highlights: • Macrophytobenthic plants were applied as a bioindicators for heavy metals pollution assessment. • The environmental quality standards for Pb, Cd, Ni, Hg in macrophytobenthic plants were evaluated. • The marine environment status assessment method based on bioindicators was proposed.

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

International Nuclear Information System (INIS)

1995-08-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner's Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section

Angra nuclear plant - environmental control program; Usina Nuclear de Angra: programa de controle ambiental

Energy Technology Data Exchange (ETDEWEB)

Kircher, E; Cruz, E.S. da [FURNAS, Rio de Janeiro, RJ (Brazil)

1990-12-31

The pre-operational studies, that were elaborated before the beginning of Angra I Power Plant operation, are described in particular the environmental radiological safety area till the fuel loading in the core reactor. Several aspects are included, as socio-economic survey, seismological analysis, Meteorological Program, marine biology, water cooling system, exposure measures of natural radiation, marine sediments characterization in the effluent dispersion area and Environmental Radiological Monitoring Program. The main environmental programs developed for the operational phase of the Angra I Plant are also presented, citing some considerations about the Meteorological Program, Marine Biology Control Program, Temperature and Chlorine Control in Piraquara de Fora Bay, Radiological Environmental Monitoring Program, Sanitary Effluent Control Program and Radiological Emergency Program. (C.G.C.). 2 refs.

Recent computer applications in boiling water reactor power plants

International Nuclear Information System (INIS)

Hiraga, Shoji; Joge, Toshio; Kiyokawa, Kazuhiro; Kato, Kanji; Nigawara, Seiitsu

1976-01-01

Process computers in boiling water reactor power plants have won the position of important equipments for the calculation of the core and plant performances and for data logging. Their application technique is growing larger and larger every year. Here, two systems are introduced; plant diagnostic system and computerized control panel. The plant diagnostic system consists of the part processing the signals from a plant, the operation part mainly composed of a computer to diagnose the operating conditions of each system component using input signal, and the result display (CRT or typewriter). The concept on the indications on control panels in nuclear power plants is changing from ''Plant parameters and to be indicated on panel meters as much as possible'' to ''Only the data required for operation are to be indicated.'' Thus the computerized control panel is attracting attention, in which the process computer for processing the operating information and CRT display are introduced. The experimental model of that panel comprises and operator's console and a chief watchmen's console. Its functions are dialogic data access and the automatic selection of preferential information. (Wakatsuki, Y.)

High-temperature gas-cooled reactor steam-cycle/cogeneration lead plant. Plant Protection and Instrumentation System design description

International Nuclear Information System (INIS)

1983-01-01

The Plant Protection and Instrumentation System provides plant safety system sense and command features, actuation of plant safety system execute features, preventive features which maintain safety system integrity, and safety-related instrumentation which monitors the plant and its safety systems. The primary function of the Plant Protection and Instrumentation system is to sense plant process variables to detect abnormal plant conditions and to provide input to actuation devices directly controlling equipment required to mitigate the consequences of design basis events to protect the public health and safety. The secondary functions of the Plant Protection and Instrumentation System are to provide plant preventive features, sybsystems that monitor plant safety systems status, subsystems that monitor the plant under normal operating and accident conditions, safety-related controls which allow control of reactor shutdown and cooling from a remote shutdown area

Development step toward fusion power plant and role of experimental reactor ITER

International Nuclear Information System (INIS)

Hiwatari, Ryouji; Asaoka, Yoshiyuki; Okano, Kunihiko

2005-01-01

The development of fusion energy is going into the experimental reactor stage, and the thermal energy from the fusion reaction will be generated in a plant scale through the ITER (International Thermonuclear Experimental Reactor) project. The remaining critical issue toward the realization of fusion energy is to map out the development strategy. Recently early realization approach as for the fusion energy development is being discussed in Japan, Europe, and the United States. This approach implies that the devices for a Demo reactor and a proto-type reactor as seen in the fast breeder reactor are combined into a single device in order to advance the fusion energy development. On the other hand, a clear development road map for fusion energy hasn't been suggested yet, and whether that early realization approach is feasible or not is still ambiguous. In order to realize the fusion energy as an user-friendly energy system, the suggestion of the development missions and the road map from the user-side point of view is instructive not only to Japanese but also to other country's development policy after the ITER project. In this report, first of all, the development missions from the user's point of view have been structured. Second, the development target required to demonstrate net electric generation and to introduce the fusion energy into the market is investigated, respectively. This investigation reveals that the completion of the ITER reference operation gives the outlook toward the demonstration of net electric generation and that the completion of the ITER advanced operation gives the possibility to introduce the fusion energy into the market. At last, the electric demonstration power plant Demo-CREST and the commercial power plant CREST are proposed to construct the development road map for fusion energy. (author)

Nuclear power plant monitoring method by neural network and its application to actual nuclear reactor

International Nuclear Information System (INIS)

Nabeshima, Kunihiko; Suzuki, Katsuo; Shinohara, Yoshikuni; Tuerkcan, E.

1995-11-01

In this paper, the anomaly detection method for nuclear power plant monitoring and its program are described by using a neural network approach, which is based on the deviation between measured signals and output signals of neural network model. The neural network used in this study has three layered auto-associative network with 12 input/output, and backpropagation algorithm is adopted for learning. Furthermore, to obtain better dynamical model of the reactor plant, a new learning technique was developed in which the learning process of the present neural network is divided into initial and adaptive learning modes. The test results at the actual nuclear reactor shows that the neural network plant monitoring system is successfull in detecting in real-time the symptom of small anomaly over a wide power range including reactor start-up, shut-down and stationary operation. (author)

Nuclear Security for Floating Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Skiba, James M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scherer, Carolynn P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-10-13

Recently there has been a lot of interest in small modular reactors. A specific type of these small modular reactors (SMR,) are marine based power plants called floating nuclear power plants (FNPP). These FNPPs are typically built by countries with extensive knowledge of nuclear energy, such as Russia, France, China and the US. These FNPPs are built in one country and then sent to countries in need of power and/or seawater desalination. Fifteen countries have expressed interest in acquiring such power stations. Some designs for such power stations are briefly summarized. Several different avenues for cooperation in FNPP technology are proposed, including IAEA nuclear security (i.e. safeguards), multilateral or bilateral agreements, and working with Russian design that incorporates nuclear safeguards for IAEA inspections in non-nuclear weapons states

Halden Boiling Water Reactor. Plant Performance and Heavy-Water Management

Energy Technology Data Exchange (ETDEWEB)

Aas, S.; Jamne, E.; Wullum, T.; Fjellestad, K. [Institutt for Atomenergi, OECD Halden Reactor Project, Halden (Norway)

1968-04-15

The Halden boiling heavy-water reactor, designed and built by the Norwegian Institutt for Atomenergi, has since June 1958 been operated as an international project. On its second charge the reactor was operated at power levels up to 25 MW and most of the time at a pressure of 28.5 kg/cm{sup 2}. During the period from July 1964 to December 1966 the plant availability was close to 64% including shutdowns because of test fuel failures and loading/unloading of fuel. Disregarding such stops, the availability was close to 90%. The average burnup of the core is about 6200 MWd/t UO{sub 2} : the most highly exposed elements have reached 10000 MWd/t UO{sub 2}. The transition temperature of the reactor tank has been followed closely. The results of the surveillance programme and the implication on the reactor operation are discussed. The reactor is located in a cave in a rock. Some experiences with such a containment are given. To locate failed test-fuel elements a fuel failure location system has been installed. A fission gas collection system has saved valuable reactor time during clean-up of the reactor system following test fuel failures. Apart from one incident with two of the control stations, the plant control and instrumentation systems have functioned satisfactorily. Two incidents with losses of 150 and 200 kg of heavy water have occurred. However, after improved methods for leakage detection had been developed, the losses have been kept better than 50 g/h . Since April 1962 the isotopic purity of the heavy water (14 t) has decreased from 99.75 to 99.62%. The tritium concentration is now slightly above 700 {mu}C/cm{sup 3}. This activity level has not created any serious operational or maintenance problems. An extensive series of water chemistry experiments has been performed to study the influence of various operating parameters on radiolytic gas formation. The main results of these experiments will be reported. Different materials such as mild steel, ferritic steel

Study for Safeguards Challenges to the Most Probably First Indonesian Future Power Plant of the Pebble Bed Modular Reactor

International Nuclear Information System (INIS)

Susilowati, E.

2015-01-01

In the near future Indonesia, the fourth most populous country, plans to build a small size power plant most probably a Pebble Bed Modular Reactor PBMR. This first nuclear power plant (NPP) is aimed to provide clear picture to the society in regard to performance and safety of nuclear power plant operation. Selection to the PBMR based on several factor including the combination of small size of the reactor and type of fuel allowing the use of passive safety systems, resulting in essential advantages in nuclear plant design and less dependence on plant operators for safety. In the light of safeguards perspective this typical reactor is also quite difference with previous light water reactor (LWR) design. From the fact that there are a small size large number of elements present in the reactor produced without individual serial numbers combine to on-line refueling same as the CANDU reactor, enforcing a new challenge to safeguards approach for this typical reactor. This paper discusses a bunch of safeguards measures have to be prepared by facility operator to support successfully international nuclear material and facility verification including elements of design relevant to safeguards need to be accomplished in consultation to the regulatory body, supplier or designer and the Agency/IAEA such as nuclear material balance area and key measurement point; possible diversion scenarios and safeguards strategy; and design features relevant to the IAEA equipment have to be installed at the reactor facility. It is deemed that result of discussion will alleviate and support the Agency approaching safeguards measure that may be applied to the purpose Indonesian first power plant of PBMR construction and operation. (author)

The computer code system for reactor radiation shielding in design of nuclear power plant

International Nuclear Information System (INIS)

Li Chunhuai; Fu Shouxin; Liu Guilian

1995-01-01

The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

Feasibility of a single-purpose reactor plant for district heating in Finland

International Nuclear Information System (INIS)

Tarjanne, R.; Vuori, S.; Eerikaeinen, L.; Saukkoriipi, L.

A feasibility study of a single-purpose reactor for district heating is presented. The reactor chosen is of an ordinary pressurized water reactor type with a thermal output of 100 to 200 MW. Primary circuit steam generators employed in ordinary PWR's are replaced by water-water heat exchangers. For safety reasons an intermediate circuit separates the primary from the network water. The conditions of the district heating systems in Finland were taken into account, which led to the choice of an average temperature of 160 0 C for the reactor coolant and a pressure of 13.5 bar. This, coupled with minimal control requirements helped design a considerably simple reactor plant. On condition, the reactor satisfies the basic heat demand in a district heating system, the effective annual full-power operation time of the heating reactor is from 5000 h to 7000 h. Economic comparisons indicated that the heating reactor may be competitive if the operation period is of this order. As the reactor has to be sited near the heat consumption area for reasons of economy, the safety aspects are of major importance and may in themselves preclude the realization of the heating idea. (author)

Chemical aspects of gadolinium nitrate as soluble nuclear poison in Savannah River Plant reactors

International Nuclear Information System (INIS)

Baumann, E.W.

1978-01-01

The aqueous solution chemistry of gadolinium nitrate was studied to identify conditions that interfere with successful cleanup of gadolinium in Savannah River Plant reactor systems. Injecting a gadolinium nitrate solution into the D 2 O coolant-moderator constitutes a supplementary mode of reactor shutdown. The resulting approximately 0.001M gadolinium nitrate solution is then deionized by recirculation through mixed-bed ion exchange resins before reactor operation is resumed

Dynamic simulation of a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Shinaishin, M.A.M.

1976-01-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is dealt with. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. In addition to the usual assumption of lumped parameters, uniform heat transfer and point kinetics (prompt jump) have been the main approximations in this and other simulators (see below). Two different transport-delay models have also been installed in all simulators. The simulators were constructed using the DARE-P System, developed by the Electrical Engineering Department at the University of Arizona

DCNS invents the immersed civil modular reactor

International Nuclear Information System (INIS)

Guilhem, Jean

2013-01-01

SMRs (Small and Modular Reactors) are a response to networks smaller than 10 GW. They are under study in USA, Russia and China, and a Korean one is already certified. In France, DCNS proposes Flexblue, a compact modular 160 MWe reactor, completely immersed 60 to 100 m under the sea level, few kilometres off the coasts. Almost invisible and protected by the marine environment, it will benefit from an infinite cold source which will ensure a high safety level. It is designed to produce 1 MWh at less than 100 euros. It will be retrieved onshore for its dismantling at the end of its service life. Its operation is said to be neutral for surrounding ecosystems with a tritium release more than 90 per cent less than that of onshore power plants

Brazilian nuclear power plants decommissioning plan for a multiple reactor site

Energy Technology Data Exchange (ETDEWEB)

Monteiro, Deiglys B.; Moreira, Joao M.L.; Maiorino, Jose R., E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas. Programa de Pos-Graduacao em Energia e Engenharia da Energia

2015-07-01

Actually, Brazil has two operating Nuclear Power Plants and a third one under construction, all at Central Nuclear Almirante Alvaro Alberto - CNAAA. To comply with regulatory aspects the power plants operator, Eletronuclear, must present to Brazilian Nuclear Regulatory Agency, CNEN, a decommissioning plan. Brazilian experience with decommissioning is limited because none of any nuclear reactor at the country was decommissioned. In literature, decommissioning process is well described despite few nuclear power reactors have been decommissioned around the world. Some different approach is desirable for multiple reactors sites, case of CNAAA site. During the decommissioning, a great amount of wastes will be produced and have to be properly managed. Particularly, the construction of Auxiliary Services on the site could be a good choice due to the possibility of reducing costs. The present work intends to present to the Eletronuclear some aspects of the decommissioning concept and decommissioning management, storage and disposal de wastes, based on the available literature, regulatory standards of CNEN and international experience as well as to suggest some solutions to be implemented at CNAAA site before starts the decommissioning project in order to maximize the benefits. (author)

Brazilian nuclear power plants decommissioning plan for a multiple reactor site

International Nuclear Information System (INIS)

Monteiro, Deiglys B.; Moreira, Joao M.L.; Maiorino, Jose R.

2015-01-01

Actually, Brazil has two operating Nuclear Power Plants and a third one under construction, all at Central Nuclear Almirante Alvaro Alberto - CNAAA. To comply with regulatory aspects the power plants operator, Eletronuclear, must present to Brazilian Nuclear Regulatory Agency, CNEN, a decommissioning plan. Brazilian experience with decommissioning is limited because none of any nuclear reactor at the country was decommissioned. In literature, decommissioning process is well described despite few nuclear power reactors have been decommissioned around the world. Some different approach is desirable for multiple reactors sites, case of CNAAA site. During the decommissioning, a great amount of wastes will be produced and have to be properly managed. Particularly, the construction of Auxiliary Services on the site could be a good choice due to the possibility of reducing costs. The present work intends to present to the Eletronuclear some aspects of the decommissioning concept and decommissioning management, storage and disposal de wastes, based on the available literature, regulatory standards of CNEN and international experience as well as to suggest some solutions to be implemented at CNAAA site before starts the decommissioning project in order to maximize the benefits. (author)

Power Reactor Fuel Reprocessing Plant-2, Tarapur: a benchmark in Indian PHWR spent fuel reprocessing

International Nuclear Information System (INIS)

Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

2017-01-01

Power Reactor Fuel Reprocessing Plant-2 (PREFRE-2) is latest operating spent nuclear fuel reprocessing plant in India. This plant has improved design based on latest technology and feedback provided by the earlier plants. The design of PREFRE-2 plant is in five cycles of solvent extraction using TBP as extractant. The plant is commissioned in year 2011 after regulatory clearances

Reactor instrumentation and control in nuclear power plants in Germany

International Nuclear Information System (INIS)

Aleite, W.

1993-01-01

The pertinent legislation, guidelines and standards of importance for nuclear power plant construction as well as the relevant committees in Germany are covered. The impact of international developments on the German regulatory scene is mentioned. A series of 15 data sheets on reactor control, followed by 5 data sheets on instrumentation and control in nuclear power plants, which were drawn up for German plants, are compared and commented in some detail. Digitalization of instrumentation and control systems continues apace. To illustrate the results that can be achieved with a digitalized information system, a picture series that documents a plant test of behavior on simulated steam generator tube rupture is elaborately commented. An outlook on backfitting and upgrading applications concludes this paper. (orig.) [de

Burst protected nuclear reactor plant with PWR

International Nuclear Information System (INIS)

Harand, E.; Michel, E.

1978-01-01

In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

Limiting factor analysis of high availability nuclear plants (boiling water reactors). Final report

International Nuclear Information System (INIS)

Frederick, L.G.; Brady, R.M.; Shor, S.W.W.; McCusker, J.T.; Alden, W.M.; Kovacs, S.

1979-08-01

The pertinent results are presented of a 16-month study conducted for Electric Power Research Institute by General Electric Company, Bechtel Power Corporation, and Philadelphia Electric Company. The study centered around the Peach Bottom 2 Atomic Power Station, but also included limited study of operations at 20 additional operating boiling water reactors. The purpose of the study was to identify and evaluate key factors limiting plant availability, and to identify potential improvements for eliminating or alleviating those limitations. The key limiting factors were found to be refueling activities; activities related to the reactor fuel; reactor scrams; activities related to 20 operating systems or major components; delays due to radiation, turbid water during refueling operations, facilities/working conditions, and dirt/foreign material; and general maintenance/repair of valves and piping. Existing programs to reduce the effect on plant unavailability are identified, and suggestions for further action are made

Consequences of severe radioactive releases to Nordic Marine environment

Energy Technology Data Exchange (ETDEWEB)

Iosjpe, M. [Norwegian Radiation Protection Authority (NRPA) (Norway); Isaksson, M. [Univ. of Gothenburg (Sweden); Joensen, H.P. [Froskaparsetur Foeroya. Faroe Islands, Torshavn (Denmark); Lahtinen, J. [Radiation and Nuclear Safety Authority (Finland); Logemann, K. [Univ. of Iceland (Iceland); Palsson, S.E. [Geislavarnir Rikisins (Iceland); Roos, P. [Technical Univ. of Denmark. Risoe DTU, Roskilde (Denmark); Suolanen, V. [Technical Research Centre of Finland (Finland)

2013-02-15

In the report, consequences of hypothetical severe nuclear accidents releases to Nordic marine environment are preliminary considered. The considered marine area comprises the Baltic Sea (Sweden, Denmark, Finland) and the North Atlantic (Iceland, Faroes, Norway) areas. The hypothetical severe nuclear accidents can be related to nuclear power plants, nuclear powered submarines or ice-breakers. Quite comprehensive survey on radioactive source terms of extremely severe nuclear power and submarine accidents has been done. This enables to estimate more realistically possible radioactive releases of various elements and nuclides to marine environment. One recent reference is of course the Fukushima accident and estimated releases there. The marine flows and dilution circumstances around the Nordic nuclear power plants and in the Baltic Sea area in general, has been studied. Respectively marine flows related to Iceland and Faroes coasts are considered with measured data and with preliminary 3D-model simulations. The substantial depth of sea water in the North Atlantic affect vertical concentration profiles to some extent. At Icelandic or Faroese waters, a potential submarine accident would likely occur in a well defined water mass, and radioactivity from the accident would be detected and spread with the flow regime of the water mass in the world ocean. Based on hypothetical severe accidents scenarios, preliminary consequence calculations has been done. It should be emphasised that the considered severe accident cases, considered in this study, do not directly attach any specific Nordic nuclear power plant or any specific submarine type. The considered radioactive releases will, however, provide specified references for more extensive consideration of environmental consequences of severe - or minor - radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MW{sub th} reactor size were used. Based on source term analyses, the

Consequences of severe radioactive releases to Nordic Marine environment

International Nuclear Information System (INIS)

Iosjpe, M.; Isaksson, M.; Joensen, H.P.; Lahtinen, J.; Logemann, K.; Palsson, S.E.; Roos, P.; Suolanen, V.

2013-02-01

In the report, consequences of hypothetical severe nuclear accidents releases to Nordic marine environment are preliminary considered. The considered marine area comprises the Baltic Sea (Sweden, Denmark, Finland) and the North Atlantic (Iceland, Faroes, Norway) areas. The hypothetical severe nuclear accidents can be related to nuclear power plants, nuclear powered submarines or ice-breakers. Quite comprehensive survey on radioactive source terms of extremely severe nuclear power and submarine accidents has been done. This enables to estimate more realistically possible radioactive releases of various elements and nuclides to marine environment. One recent reference is of course the Fukushima accident and estimated releases there. The marine flows and dilution circumstances around the Nordic nuclear power plants and in the Baltic Sea area in general, has been studied. Respectively marine flows related to Iceland and Faroes coasts are considered with measured data and with preliminary 3D-model simulations. The substantial depth of sea water in the North Atlantic affect vertical concentration profiles to some extent. At Icelandic or Faroese waters, a potential submarine accident would likely occur in a well defined water mass, and radioactivity from the accident would be detected and spread with the flow regime of the water mass in the world ocean. Based on hypothetical severe accidents scenarios, preliminary consequence calculations has been done. It should be emphasised that the considered severe accident cases, considered in this study, do not directly attach any specific Nordic nuclear power plant or any specific submarine type. The considered radioactive releases will, however, provide specified references for more extensive consideration of environmental consequences of severe - or minor - radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MW th reactor size were used. Based on source term analyses, the chosen

Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

Energy Technology Data Exchange (ETDEWEB)

Wood, RT

2004-09-27

This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

Concentration factors of radionuclides in the marine organisms

International Nuclear Information System (INIS)

1996-03-01

Parameters related to the bioconcentration of radionuclides in the marine were shown by 'Assessment and guideline to the target value of dose in the environment of the power light water reactor facilities' (Nuclear Safety Commission), but the guideline data did not contain Ru and Ce relating to the reprocessing plant. So that more new data than these of 'Technical Reports Series No. 247' (published by IAEA in 1985) were mainly collected. Especially the data of nuclides with poor data of concentration factors (CF) and natural radionuclides (Po-210, Pb-210) were gathered. These data were pigeonholed and many tables (element, kinds of organisms, experimental methods) were made by separating the general remarks from the original experimental reports. The contents of this report are given as under, history of concentration factor (CF), determination method of CF, CF calculation method, calculation models related to CF, tables of metabolic parameters, tables of CF, the present conditions of studies for uptake of radionuclides with long half-life into the marine organisms, CF abstract tables and trial calculation of human exposure by eating the marine organisms. (S.Y.)

Metabolic and environmental aspects of fusion reactor activation products: niobium

Energy Technology Data Exchange (ETDEWEB)

Easterly, C.E.; Shank, K.E.

1977-11-01

A summary of the metabolic and environmental aspects of niobium is presented. The toxicological symptoms from exposure to niobium are given, along with lethal concentration values for acute and chronic exposures. Existing human data are presented; animal uptake and retention data are analyzed for various routes of administration. Recommended metabolic values are also presented along with comments concerning their use and appropriateness. The natural distribution of niobium is given for freshwater, seawater, and the biosphere. Concentration factors and retention of /sup 95/Nb in the environment are discussed with reference to: plant retention via leaf absorption; plant retention via root uptake; uptake in terrestrial animals from plants; uptake in freshwater organisms; uptake in marine organisms; and movement in soil. Conclusions are drawn regarding needs for future work in these areas. This review was undertaken because niobium is expected to be a key metal in the development of commercial fusion reactors. It is recognized that niobium will likely not be used in the first generation reactors as a structural material but will appear as an alloy in such materials as superconducting wire.

Metabolic and environmental aspects of fusion reactor activation products: niobium

International Nuclear Information System (INIS)

Easterly, C.E.; Shank, K.E.

1977-11-01

A summary of the metabolic and environmental aspects of niobium is presented. The toxicological symptoms from exposure to niobium are given, along with lethal concentration values for acute and chronic exposures. Existing human data are presented; animal uptake and retention data are analyzed for various routes of administration. Recommended metabolic values are also presented along with comments concerning their use and appropriateness. The natural distribution of niobium is given for freshwater, seawater, and the biosphere. Concentration factors and retention of 95 Nb in the environment are discussed with reference to: plant retention via leaf absorption; plant retention via root uptake; uptake in terrestrial animals from plants; uptake in freshwater organisms; uptake in marine organisms; and movement in soil. Conclusions are drawn regarding needs for future work in these areas. This review was undertaken because niobium is expected to be a key metal in the development of commercial fusion reactors. It is recognized that niobium will likely not be used in the first generation reactors as a structural material but will appear as an alloy in such materials as superconducting wire

A multi-stage-flash desalination plant of relative small performance with an integrated pressurized water reactor as a nuclear heat source

International Nuclear Information System (INIS)

Peltzer, M.; Petersen, G.

1976-01-01

In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination-plants with a performance in the range of 10,000 to 80,000 m 3 /d heated by a nuclear reactor are investigated. The reactor concept is similar to the integrated pressurized water reactor (IPWR) of the nuclear ship OTTO HAHN. The calculated costs of the desalinated water show, that due to the fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (orig.) [de

Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

1993-01-01

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW t (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment

Selection of power plant elements for future reactor space electric power systems

International Nuclear Information System (INIS)

Buden, D.; Bennett, G.A.; Copper, K.

1979-09-01

Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected

United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

International Nuclear Information System (INIS)

1997-06-01

This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

Ecological state of the Rance marine basin after 30 years of functioning of the tidal power plant

International Nuclear Information System (INIS)

Retiere, Ch.; Desroy, N.; Bonnot-Courtois, C.; Le Mao, P.

1997-01-01

The Rance basin on the Brittany coast (France) is the unique site in the world where the long term ecological impacts of the functioning of a tidal power plant can be evaluated. Two aspects are distinguished: the consequences due to the building the plant and those due to the functioning of the plant. During the building of the barrage which lasted 3 years, the estuary was isolated from the sea and led to an increase of sedimentation and organic matter and to important variations of the water salinity. Thus the whole marine flora and fauna disappeared. Today, the operation of the plant has changed the rhythm and amplitude of the tide inside the basin. Ten to fifteen years were necessary for the marine fauna and flora to recover a new equilibrium inside the basin after the plant was built. The functioning of the new ecosystem is governed by the relationship between species independently of the initial perturbation and its equilibrium is based on the regular functioning of the plant. (J.S.)

An endothermic chemical process facility coupled to a high temperature reactor. Part II: Transient simulation of accident scenarios within the chemical plant

International Nuclear Information System (INIS)

Brown, Nicholas R.; Revankar, Shripad T.

2012-01-01

Highlights: ► Seven quantitative transient case studies were analyzed in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Positive power excursion in the nuclear reactor were found for helium-inlet overcoolings. ► In all cases studied the maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Transient study of the operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven quantitative transient case studies are analyzed. The case studies consist of: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. Various parametric studies based on the magnitude of the events were also performed. The only chemical plant initiated events that caused a positive power excursion in the nuclear reactor were helium-inlet overcoolings due to process holding tank failures or reaction chamber ruptures. Even for a severe sustained overcooling, the calculated maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. The qualitative basis for the case studies and the analysis models are summarized in part I of this paper.

Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

International Nuclear Information System (INIS)

Colas, J.; Saudray, D.; Coste, J.A.; Roux, J.P.; Jouan, A.

1987-01-01

The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

Outline of construction planning on No. 2 Reactor of the Shika Nuclear Power Plant

International Nuclear Information System (INIS)

Nakagawa, Tetsuro; Kadoki, Shuichi; Kubo, Tetsuji

1999-01-01

The Hokuriku Electric Co., Ltd. carries out the expansion of the Shika Nuclear Power Plant No.2 (ABWR) to start its in March 2006. It is situated in north neighboring side of No. 1 reactor under operation at present, and its main buildings are planned to position a reactor building at mountain side and a turbine building at sea side as well as those in the No. 1 reactor. And, cooling water for steam condenser was taken in from an intake opening built at north side of the lifting space situated at the front of the power plant, and discharged into seawater from a flashing opening positioned about 600 m offing. Here were described on outline of main civil engineering such as base excavation engineering, concrete caisson production, oceanic establishment engineering, and facility for steam condenser, and characteristics of the engineering. (G.K.)

Manning designs for nuclear district-heating plant (NDHP) with RUTA-type reactor

International Nuclear Information System (INIS)

Gerasimova, V.S.; Mikhan, V.I.; Romenkov, A.A.

2001-01-01

RUTA-type reactor is a water cooled water-moderated pool-type reactor with an atmospheric pressure air medium. The reactor has been designed for heating and hot water supply. Nuclear district heating plant (NDHP) with RUTA-type reactor facility has been designed with a three circuit layout. Primary circuit components are arranged integrally in the reactor vessel. Natural coolant circulation mode is used in the primary circuit. A peculiarity of RUTA-based NDHP as engineered system is a smooth nature of its running slow variation of the parameters at transients. Necessary automation with application of computer equipment will be provided for control and monitoring of heat production process at NDHP. Under developing RUTA-based NDHP it is foreseen that operating staff performs control and monitoring of heat generation process and heat output to consumers as well as current maintenance of NDHP components. All other works associated with NDHP operation should be fulfilled by extraneous personnel. In so doing the participation of operating staff is also possible. (author)

Nuclear power desalinating complex with IRIS reactor plant and Russian distillation desalinating unit

International Nuclear Information System (INIS)

Kostin, V. I.; Panov, Yu.K.; Polunichev, V. I.; Fateev, S. A.; Gureeva, L. V.

2004-01-01

This paper has been prepared as a result of Russian activities on the development of nuclear power desalinating complex (NPDC) with the IRIS reactor plant (RP). The purpose of the activities was to develop the conceptual design of power desalinating complex (PDC) and to evaluate technical and economical indices, commercial attractiveness and economical efficiency of PDC based on an IRIS RP with distillation desalinating plants. The paper presents the main results of studies as applied to dual-purpose PDC based on IRIS RP with different types of desalinating plants, namely: characteristics of nuclear power desalinating complex based on IRIS reactor plant using Russian distillation desalinating technologies; prospective options of interface circuits of the IRIS RP with desalinating plants; evaluations of NPDC with IRIS RP output based on selected desalinating technologies for water and electric power supplied to the grid; cost of water generated by NPDC for selected interface circuits made by the IAEA DEEP code as well as by the Russian TEO-INVEST code; cost evaluation results for desalinated water of PDC operating on fossil fuel and conditions for competitiveness of the nuclear PDC based on IRIS RP compared with analog desalinating complexes operating on fossil fuel.(author)

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1989-10-01

Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule duration, and to simplify design. This document discusses construction approaches. 77 refs., 5 figs., 6 tabs

Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

International Nuclear Information System (INIS)

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%

SWR 1000: The new boiling water reactor power plant concept

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

Potential of Electric Power Production from Microbial Fuel Cell (MFC) in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

Science.gov (United States)

Zaman, Badrus; Wardhana, Irawan Wisnu

2018-02-01

Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media). Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day) operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

A small floating seawater desalination plant using a nuclear heating reactor coupled with the MED process

International Nuclear Information System (INIS)

Dong Duo; Wu Shaorong; Zhang Dafang; Wu Zongxin

1997-01-01

A small floating seawater desalination plant using a nuclear heating reactor coupled with a multi-effect distillation (MED) process was designed by the Institute of Nuclear Energy Technology, Tsinghua University of China. It was intended to supply potable water to remove coastal areas or islands where both fresh water and energy are severely lacking, and also to serve as a demonstration and training facility. The design of a small floating plant coupled two proven technologies in the cogeneration mode: a nuclear heating reactor (NHR-10), with inherent, passive safety features based on NHR-5 experience, and a low temperature MED process. The secondary loop was designed as a safety barrier between the primary loop and the steam loop. With a 10 MW(th) heating reactor, the floating plant could provide 4,000 m 3 /d of potable water and 750 kW of electricity. The design concept and parameters, safety features, coupling scheme and floating plant layout are presented in the paper. (author). 3 refs, 4 figs, 3 tabs

Evaluation of marine algae as a source of biogas in a two-stage anaerobic reactor system

International Nuclear Information System (INIS)

Vergara-Fernandez, Alberto; Vargas, Gisela; Alarcon, Nelson; Velasco, Antonio

2008-01-01

The marine algae are considered an important biomass source; however, their utilization as energy source is still low around the world. The technical feasibility of marine algae utilization as a source of renewable energy was studied to laboratory scale. The anaerobic digestion of Macrocystis pyrifera, Durvillea antarctica and their blend 1:1 (w/w) was evaluated in a two-phase anaerobic digestion system, which consisted of an anaerobic sequencing batch reactor (ASBR) and an upflow anaerobic filter (UAF). The results show that 70% of the total biogas produced in the system was generated in the UAF, and both algae species have similar biogas productions of 180.4(±1.5) mL g -1 dry algae d -1 , with a methane concentration around 65%. The same methane content was observed in biogas yield of algae blend; however, a lower biogas yield was obtained. In conclusion, either algae species or their blend can be utilized to produce methane gas in a two-phase digestion system

Evaluation of marine algae as a source of biogas in a two-stage anaerobic reactor system

Energy Technology Data Exchange (ETDEWEB)

Vergara-Fernandez, Alberto; Vargas, Gisela [Escuela de Ingenieria Ambiental, Facultad de Ingenieria, Universidad Catolica de Temuco, Manuel Montt 56, Casilla 15-D, Temuco (Chile); Alarcon, Nelson [Departamento de Ingenieria Quimica, Facultad de Ingenieria y Ciencias Geologicas, Universidad Catolica del Norte (Chile); Velasco, Antonio [Centro Nacional de Investigacion y Capacitacion Ambiental del Instituto Nacional de Ecologia (CENICA-INE), Av. San Rafael Atlixco 186, Col. Vicentina, Del. Iztapalapa, 09340, Mexico, DF (Mexico)

2008-04-15

The marine algae are considered an important biomass source; however, their utilization as energy source is still low around the world. The technical feasibility of marine algae utilization as a source of renewable energy was studied to laboratory scale. The anaerobic digestion of Macrocystis pyrifera, Durvillea antarctica and their blend 1:1 (w/w) was evaluated in a two-phase anaerobic digestion system, which consisted of an anaerobic sequencing batch reactor (ASBR) and an upflow anaerobic filter (UAF). The results show that 70% of the total biogas produced in the system was generated in the UAF, and both algae species have similar biogas productions of 180.4({+-}1.5) mL g{sup -1} dry algae d{sup -1}, with a methane concentration around 65%. The same methane content was observed in biogas yield of algae blend; however, a lower biogas yield was obtained. In conclusion, either algae species or their blend can be utilized to produce methane gas in a two-phase digestion system. (author)

Marine ecology

International Nuclear Information System (INIS)

Anon.

1977-01-01

Studies on marine ecology included marine pollution; distribution patterns of Pu and Am in the marine waters, sediments, and organisms of Bikini Atoll and the influence of physical, chemical, and biological factors on their movements through marine biogeochemical systems; transfer and dispersion of organic pollutants from an oil refinery through coastal waters; transfer of particulate pollutants, including sediments dispersed during construction of offshore power plants; and raft culture of the mangrove oysters

14th Biennial conference on reactor operating experience plant operations: The human element

International Nuclear Information System (INIS)

Anon.

1989-01-01

Separate abstracts were prepared for the papers presented in the following areas of interest: enhancing operator performance; structured approaches to maintenance standards and reliability-centered maintenance; human issues in plant operations and management; test, research, and training reactor utilization; methods and applications of root-cause analysis; emergency operating procedure enhancement programs; test, research, and training reactor upgrades; valve maintenance and diagnostics; recent operating experiences; and current maintenance issues

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1987-12-01

Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses successes and problems in construction. 49 refs., 16 figs., 8 tabs

Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide (Spanish Edition)

International Nuclear Information System (INIS)

2010-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.

Biodenitrification of gaseous diffusion plant aqueous wastes: stirred bed reactor

International Nuclear Information System (INIS)

Holland, M.E.

1980-01-01

Approximately 30 kilograms of nitrates per day are discarded in the raffinates (acid wastes) of the Portsmouth Gaseous Diffusion Plant's X-705 Uranium Recovery and Decontamination Facility. A biodenitrification process employing continuous-flow, stirred-bed reactors has been successfully used to remove nitrates from similar acid wastes at the Oak Ridge Y-12 Plant. Laboratory studies have been made at Portsmouth to characterize the X-705 raffinates and to test the stirred-bed biodenitrification process on such raffinates. Raffinates which had been previously characterized were pumped through continuous-flow, stirred-bed, laboratory-scale reactors. Tests were conducted over a period of 146 days and involved variations in composition, mixing requirements, and the fate of several metal ions in the raffinates. Tests results show that 20 weight percent nitrates were reduced to a target nitrate effluent concentration of 100 μg/ml with a 99.64 percent efficiency. However, the average denitrification rate achieved was only 33% of that demonstrated with the Y-12 stirred-bed system. These low rates were probably due to the toxic effects of heavy metal ions on the denitrifying bacteria. Also, most of the uranium in the raffinate feed remained in the biomass and calcite, which collected in the reactor. This could cause criticality problems. For these reasons, it was decided not to make use of the stirred-bed bioreactor at Portsmouth. Instead, the biodenitrification installation now planned will use fluidized bed columns whose performance will be the subject of a subsequent report

Methodologies for assessment of power plant ecological effects in the marine environment

International Nuclear Information System (INIS)

Adams, S.M.

1978-01-01

Various types of models or methodologies relevant to the assessment of entrainment, thermal, and impingement impacts of power plant operation in the marine environment are presented. The majority of methodologies available for assessing power plant effects are focused at the organism or population level. The most widely applied approaches for estimating entrainment effects on fish populations are the equivalent adult and trophic-conversion methodologies. Current methods to predict the number of fish and distribution of species impinged consider physical factors of the environment but not the biological or behavorial characteristics of fish. With proper validation, ecosystem-level models that consider aggregate responses of biological components to stress may prove to be a viable approach for investigating power plant ecological effects

Liquid radioactive waste processing system for pressurized water reactor plants

International Nuclear Information System (INIS)

Anon.

1976-01-01

This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

Potential of Electric Power Production from Microbial Fuel Cell (MFC in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

Directory of Open Access Journals (Sweden)

Zaman Badrus

2018-01-01

Full Text Available Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media. Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-01-01

Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

Seismic monitoring of the Creys-Malville plant - Problems raised by the seismic behaviour of a fast breeder reactor

International Nuclear Information System (INIS)

Descleve, P.; Barrau, P.

1988-01-01

CREYS-MALVILLE reached full power in December 1986 and is presently the largest sodium cooled reactor in operation. Well established procedures of safety evaluation have been used for the design but for a large size reactor special attention must be paid to the effects of seismic disturbances. This paper describes the seismic protection and monitoring system of the plant, the core behaviour which is specific to fast reactors and the test performed to verify the analyses. Finally the seismic impact on the construction can be established as an indication for future plants. (author)

An endothermic chemical process facility coupled to a high temperature reactor. Part I: Proposed accident scenarios within the chemical plant

International Nuclear Information System (INIS)

Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

2012-01-01

Highlights: ► The paper identifies possible transient and accident scenarios in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Key accidents scenarios were investigated through qualitative reasoning. ► The accidents were found to constitute loss of heat sink event for the nuclear reactor. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Quantitative study of the possible operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven unique case studies are proposed based on a thorough review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. The qualitative parameters of each case study are outlined as well as the basis in literature. A previously published modeling scheme is described and adapted for application as a simulation platform for these transient events. The results of the quantitative case studies are described within part II of this paper.

Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

Davis, G.A.

1992-01-01

Since 1985, ABB Combustion Engineering Nuclear Power (CENP) and Duke Engineering ampersand Services, Inc. (DE ampersand S) have been developing the next generation of pressurized water reactor (PWR) plant for worldwide deployment. The goal is to make available a pre-licensed, standardized plant design that can satisfy the need for a reliable and economic supply of electricity for residential, commercial and industrial use. To ensure that such a design is available when needed, it must be based on proven technology and established licensing criteria. These requirements dictate development of nuclear technology that is advanced, yet evolutionary in nature. This has been achieved with the System 80+ Standard Plant Design

Nuclear power plant control room task analysis. Pilot study for pressurized water reactors

International Nuclear Information System (INIS)

Barks, D.B.; Kozinsky, E.J.; Eckel, S.

1982-05-01

The purposes of this nuclear plant task analysis pilot study: to demonstrate the use of task analysis techniques on selected abnormal or emergency operation events in a nuclear power plant; to evaluate the use of simulator data obtained from an automated Performance Measurement System to supplement and validate data obtained by traditional task analysis methods; and to demonstrate sample applications of task analysis data to address questions pertinent to nuclear power plant operational safety: control room layout, staffing and training requirements, operating procedures, interpersonal communications, and job performance aids. Five data sources were investigated to provide information for a task analysis. These sources were (1) written operating procedures (event-based); (2) interviews with subject matter experts (the control room operators); (3) videotapes of the control room operators (senior reactor operators and reactor operators) while responding to each event in a simulator; (4) walk-/talk-throughs conducted by control room operators for each event; and (5) simulator data from the PMS

Rock siting of nuclear power plants from a reactor safety standpoint. Status report October 1974

International Nuclear Information System (INIS)

1975-01-01

The aim of this study is to clearify the advantages and disadvantages of an underground nuclear power plant from a reactor safety point of view, compared to a plant above ground. Principles for the technical design of a rock sited BWR nuclear power plant is presented. Also questions of sabotage and closing down the plant at the end of the operational period are treated. (K.K.)

Westinghouse Small Modular Reactor balance of plant and supporting systems design

Energy Technology Data Exchange (ETDEWEB)

Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

Westinghouse Small Modular Reactor balance of plant and supporting systems design

International Nuclear Information System (INIS)

Memmott, M. J.; Stansbury, C.; Taylor, C.

2012-01-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

Control room conceptual design of nuclear power plant with multiple modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Jia Qianqian; Qu Ronghong; Zhang Liangju

2014-01-01

A conceptual design of the control room layout for the nuclear power plant with multiple modular high temperature gas-cooled reactors has been developed. The modular high temperature gas-cooled reactors may need to be grouped to produce as much energy as a utility demands to realize the economic efficiency. There are many differences between the multi-modular plant and the current NPPs in the control room. These differences may include the staffing level, the human-machine interface design, the operation mode, etc. The potential challenges of the human factor engineering (HFE) in the control room of the multi-modular plant are analyzed, including the operation workload of the multi-modular tasks, how to help the crew to keep situation awareness of all modules, and how to support team work, the control of shared system between modules, etc. A concept design of control room for the multi-modular plant is presented based on the design aspect of HTR-PM (High temperature gas-cooled reactor pebble bed module). HFE issues are considered in the conceptual design of control room for the multi-modular plant and some design strategies are presented. As a novel conceptual design, verifications and validations are needed, and focus of further work is sketch out. (author)

Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

Energy Technology Data Exchange (ETDEWEB)

LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

2013-05-01

The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

Methods and technologies for cost reduction in the design of water cooled reactor power plants

International Nuclear Information System (INIS)

1991-05-01

The Specialists Meeting was organized in the framework of the IAEA International Working Group on Advanced Technologies for Water-Cooled Reactors. Its purpose was to provide an international forum for review and discussion on recent results in research and development on different methods and technologies of current and advanced water-cooled reactor power plants, which can lead to reduced investment and operation, maintenance and fuel-cycle costs of the plants. 27 specialists representing 10 countries and the IAEA took part in the meeting. 10 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Marine genomics

DEFF Research Database (Denmark)

Oliveira Ribeiro, Ângela Maria; Foote, Andrew David; Kupczok, Anne

2017-01-01

Marine ecosystems occupy 71% of the surface of our planet, yet we know little about their diversity. Although the inventory of species is continually increasing, as registered by the Census of Marine Life program, only about 10% of the estimated two million marine species are known. This lag......-throughput sequencing approaches have been helping to improve our knowledge of marine biodiversity, from the rich microbial biota that forms the base of the tree of life to a wealth of plant and animal species. In this review, we present an overview of the applications of genomics to the study of marine life, from...

Chemical reactor for a PUREX reprocessing plant of 200Kg U/day capacity

International Nuclear Information System (INIS)

Oliveria Lopes, M.J. de.

1974-03-01

Dissolution of spent reactor fuels in Purex process is studied. Design of a chemical reactor for PWR elements, 3% enriched uranium dioxide with zircaloy cladding, for a 200Kg/day uranium plant is the main objective. Chop-leach process is employed and 7.5M nitric acid is used. Non-criticality was obtained by safe geometry and checked by spectrum homogeneous calculus and diffusion codes. Fuel cycle is considered and decladding and dissolution are treated more accurately

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1989-01-01

Under Contract No. AC03-86SF16565, Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses design requirements. 36 refs., 57 figs., 56 tabs

Radiological considerations in the design of Reprocessing Uranium Plant (RUP) of Fast Reactor Fuel Cycle Facility (FRFCF), Kalpakkam

International Nuclear Information System (INIS)

Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

2012-01-01

A Fast Reactor Fuel Cycle Facility (FRFCF) being planned at Indira Gandhi Centre for Atomic Research, Kalpakkam is an integrated facility with head end and back end of fuel cycle plants co-located in a single place, to meet the refuelling needs of the prototype fast breeder reactor (PFBR). Reprocessed uranium oxide plant (RUP) is one such plant in FRFCF to built to meet annual requirements of UO 2 for fabrication of fuel sub-assemblies (FSAs) and radial blanket sub-assemblies (RSAs) for PFBR. RUP receives reprocessed uranium oxide powder (U 3 O 8 ) from fast reactor fuel reprocessing plant (FRP) of FRFCF. Unlike natural uranium oxide plant, RUP has to handle reprocessed uranium oxide which is likely to have residual fission products activity in addition to traces of plutonium. As the fuel used for PFBR is recycled within these plants, formation of higher actinides in the case of plutonium and formation of higher levels of 232 U in the uranium product would be a radiological problem to be reckoned with. The paper discussed the impact of handling of multi-recycled reprocessed uranium in RUP and the radiological considerations

Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

1993-09-01

For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

Advanced Reactor Licensing: Experience with Digital I and C Technology in Evolutionary Plants

International Nuclear Information System (INIS)

Wood, RT

2004-01-01

This report presents the findings from a study of experience with digital instrumentation and controls (I and C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l and C systems and identified lessons learned. The report (1) gives an overview of the modern l and C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States

Emergency cooling system for a nuclear reactor in a closed gas turbine plant

International Nuclear Information System (INIS)

Frutschi, H.U.

1974-01-01

In undisturbed operation of the closed gas turbine plant with compressor stages, reactor, and turbine, a compressor stage driven by a separate motor is following with reduced power. The power input this way is so small that the working medium is just blown through without pressure increase. The compressor stage is connected with the reactor by means of a reactor feedback pipe with an additional cooler and with the other compressor stages by means of a recuperator in the pipe between these and the turbine. In case of emergency cooling, e.g. after the rupture of a pipe with decreasing pressure of the working medium, the feedback pipe is closed short and the additional compressor stage is brought to higher power. It serves as a coolant blower and transfers the necessary amount of working medium to the reactor. The compressor stage is controlled at a constant torque, so that the heat removal from the reactor is adapted to the conditions of the accident. (DG) [de

TMI-2 reactor vessel and balance of plant status

International Nuclear Information System (INIS)

Kuehn, G.A.

1990-01-01

In the fall of 1988 a corporate decision was made which concentrated effort on support of reactor vessel defueling and minimized activity in balance-of-plant areas. The auxiliary and fuel handling building are in a safe/stable state but final preparations for monitored storage won't be pursued until defueling and fuel shipping are complete. In addition to dispositioning fuel, the project is actively preparing for disposal of the Accident Generated Water (2.3 million gallons) by evaporation

Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

International Nuclear Information System (INIS)

McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

1988-01-01

Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR [pressurized-water-reactor] plants

International Nuclear Information System (INIS)

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1988-01-01

Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs

Siting of light-water reactor power plants in the Federal Republic of Germany

International Nuclear Information System (INIS)

Kohler, H.A.G.

1975-01-01

The nuclear power plant site requirements formulated for environment protection in Germany allow nuclear power plants to be built at any site provided these requirements are duly taken into account in preparing and monitoring the site and in the design of the proposed power plant. After a brief discussion of light water reactor power plant sites, prevailing practice in site planning, site selection criteria, licensing procedure and used criteria, rules and guidelines, this paper reports on some considerations taken into account by the expert advisers and by the licensing authorities and future site planning. (orig.) [de

Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

International Nuclear Information System (INIS)

Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

2015-01-01

At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

Revisit of analytical methods for the process and plant control analyses during reprocessing of fast reactor fuels

International Nuclear Information System (INIS)

Subba Rao, R.V.

2016-01-01

CORAL (COmpact facility for Reprocessing of Advanced fuels in Lead cell) is an experimental facility for demonstrating the reprocessing of irradiated fast reactor fuels discharged from the Fast Breeder Test Reactor (FBTR). The objective of the reprocessing plant is to achieve nuclear grade plutonium and uranium oxides with minimum process waste volumes. The process flow sheet for the reprocessing of spent Fast Reactor Fuel consists of Transport of spent fuel, Chopping, Dissolution, Feed conditioning, Solvent Extraction cycle, Partitioning Cycle and Re-conversion of Plutonium nitrate and uranium nitrate to respective oxides. The efficiency and performance of the plant to achieve desired objective depends on the analyses of various species in the different steps adopted during reprocessing of fuels. The analytical requirements in the plant can be broadly classified as 1. Process control Analyses (Analyses which effect the performance of the plant- PCA); 2. Plant control Analyses (Analyses which indicates efficiency of the plant-PLCA); 3. Nuclear Material Accounting samples (Analyses which has bearing on nuclear material accounting in the plant - NUMAC) and Quality control Analyses (Quality of the input bulk chemicals as well as products - QCA). The analytical methods selected are based on the duration of analyses, precision and accuracies required for each type analytical requirement classified earlier. The process and plant control analyses requires lower precision and accuracies as compared to NUMAC analyses, which requires very high precision accuracy. The time taken for analyses should be as lower as possible for process and plant control analyses as compared to NUMAC analyses. The analytical methods required for determining U and Pu in process and plant samples from FRFR will be different as compared to samples from TRFR (Thermal Reactor Fuel Reprocessing) due to higher Pu to U ratio in FRFR as compared TRFR and they should be such that they can be easily

Development of operator thinking model and its application to nuclear reactor plant operation system

International Nuclear Information System (INIS)

Miki, Tetsushi; Endou, Akira; Himeno, Yoshiaki

1992-01-01

At first, this paper presents the developing method of an operator thinking model and the outline of the developed model. In next, it describes the nuclear reactor plant operation system which has been developed based on this model. Finally, it has been confirmed that the method described in this paper is very effective in order to construct expert systems which replace the reactor operator's role with AI (artificial intelligence) systems. (author)

Inducer pumps for liquid metal reactor plants

International Nuclear Information System (INIS)

Jackson, E.D.

2002-01-01

Pumps proposed for liquid metal reactor plants typically use centrifugal impellers as the rotating element and are required to maintain a relatively low speed to keep the suction specific speed low enough to operate at the available net positive suction head (HPSH) and to avoid cavitation damage. These low speeds of operation require that the pump diameter increase and/or multiple stages be used to achieve the design head. This frequently results in a large, heavy, complex pump design. In addition, the low speed results in a larger drive motor size so that the resultant penalty to the plant designer is multiplied. The heavier pump can also result in further complications as, e.g., the difficulty in maintaining the first critical speed sufficiently above the pump operating range to provide margin for rotor dynamic stability. To overcome some of these disadvantages, it was proposed the use of inducer pumps for Liquid Metal Fast Breeder Reactor (LMFBR) plants. This paper discusses some of the advantages of the inducer pump and the development history of designing and testing these pumps both in water and sodium. The inducer pump is seen to be a sound concept with a strong technology base derived from the aerospace and ship propulsion industries. The superior suction performance capability of the inducer offers significant system design advantages, primarily a smaller, lighter weight, less complex pump design with resulting saving in cost. Extensive testing of these pumps has been conducted in both sodium and water to demonstrate the long-life capability with no cavitation damage occurring in those designs based on Rockwell's current design criteria. These tests have utilized multiple inspection and measurement approaches to accurately assess and identify any potential for cavitation damage, and these approaches have all concluded that no damage is occurring. Therefore, it is concluded that inducer pumps can be safely designed for long life operation in sodium with

The low-temperature water-water reactor for district heating atomic power plant (DHPP)

International Nuclear Information System (INIS)

Skvortsov, S.A.; Sokolov, I.N.; Krauze, L.V.; Nikiporetz, Yu.G.; Philimonov, Yu.V.

1977-01-01

The district heating atomic power plant in the article is distinguished by the increased reliability and safety of operation that was provided by the use of following main principles: relatively low parameters of the coolant; the intergral arrangement of equipment and accordingly the minimum branching of the reactor circuit; the natural circulation of coolant of the primary circuit in the steady-state, transient and emergency regimes of reactor operation; the considerable reserves of cold water of the primary circuit in the reactor vessel, providing the emergency cooling; the application of two shells each of which is designed for the total working pressure, the second shell is made of prestressed reinforced concrete that eliminates its brittle failure. (M.S.)

Install and operate type radiation processing plant for marine products

Energy Technology Data Exchange (ETDEWEB)

Kohli, A.K. [BARC-BTIR Complex, Mumbai (India). Dept. of Atomic Energy. Board of Radiation and Isotope Technology

2002-07-01

Marine products can be carrier of several pathogens. Radiation processing is a very useful technique that is used to eliminate pathogens and also to extend shelf life of fresh fish. For marine products three processes are involved namely: radurization to pasteurize fresh chilled fish for extending shelf life; radicidation to sanitize frozen fishery products by elimination of pathogenic microorganisms and radiation disinfestations to eliminate insects from dehydrated fishery products. The paper brings out conceptual design of a compact radiation processing plant that can cater to all the three processes. The design is different from conveyor type of designs. The design is specially configured to maintain the temperature of frozen products and overdose ratio within limits specified. The throughput depends upon the source strength, type of product, the size of box and its configuration in which these could be arranged. The design has many features, which make it a very safe, convenient and economical method for processing of such items or for that matter all the food products, which are amenable for radiation processing. (author)

Install and operate type radiation processing plant for marine products

International Nuclear Information System (INIS)

Kohli, A.K.

2002-01-01

Marine products can be carrier of several pathogens. Radiation processing is a very useful technique that is used to eliminate pathogens and also to extend shelf life of fresh fish. For marine products three processes are involved namely: radurization to pasteurize fresh chilled fish for extending shelf life; radicidation to sanitize frozen fishery products by elimination of pathogenic microorganisms and radiation disinfestations to eliminate insects from dehydrated fishery products. The paper brings out conceptual design of a compact radiation processing plant that can cater to all the three processes. The design is different from conveyor type of designs. The design is specially configured to maintain the temperature of frozen products and overdose ratio within limits specified. The throughput depends upon the source strength, type of product, the size of box and its configuration in which these could be arranged. The design has many features, which make it a very safe, convenient and economical method for processing of such items or for that matter all the food products, which are amenable for radiation processing. (author)

Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

Science.gov (United States)

Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

2010-10-01

The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

1987-01-01

A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

Nuclear power plant laundry drain treatment using membrane bio reactor

International Nuclear Information System (INIS)

Tsukamoto, Masaaki; Kohanawa, Osamu; Kinugasa, Atsushi; Ogawa, Naoki; Murogaki, Kenta

2012-01-01

In nuclear power plant, the radioactive effluent generated by washing the clothes worn in controlled area and the hand and shower water used at the controlled area are treated in laundry drain treatment system. Although various systems which treat such liquid waste preexist, the traditional treatment system has disadvantages such as high running cost and a large amount of secondary waste generation. To solve these matters, we have considered application of an activated sludge system, membrane bio reactor, which has been practically used in general industry. For nuclear power plant, the activated sludge system has been developed, tested in its adaptability and the adequacy has been proved. Some preexisting treatment systems have been replaced with this activated sludge system for the first time in a domestic nuclear power plant, and the renewal system is now in operation. The result is reported. (author)

Reliability Analysis Study of Digital Reactor Protection System in Nuclear Power Plant

International Nuclear Information System (INIS)

Guo, Xiao Ming; Liu, Tao; Tong, Jie Juan; Zhao, Jun

2011-01-01

The Digital I and C systems are believed to improve a plants safety and reliability generally. The reliability analysis of digital I and C system has become one research hotspot. Traditional fault tree method is one of means to quantify the digital I and C system reliability. Review of advanced nuclear power plant AP1000 digital protection system evaluation makes clear both the fault tree application and analysis process to the digital system reliability. One typical digital protection system special for advanced reactor has been developed, which reliability evaluation is necessary for design demonstration. The typical digital protection system construction is introduced in the paper, and the process of FMEA and fault tree application to the digital protection system reliability evaluation are described. Reliability data and bypass logic modeling are two points giving special attention in the paper. Because the factors about time sequence and feedback not exist in reactor protection system obviously, the dynamic feature of digital system is not discussed

Applicability of base-isolation R and D in non-reactor facilities to a nuclear reactor plant

International Nuclear Information System (INIS)

Seidensticker, R.W.

1989-01-01

Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. This paper reviews the research and development (R and D) programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R and D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R and D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant

Civil ships propulsion reactor plants development and operation experience, and prospects for their improvement

International Nuclear Information System (INIS)

Vasyukov, V.I.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

2000-01-01

Russia is alone country in the World possessing nuclear-powered icebreaker fleet. Phases of creation in Russia of several propulsion nuclear reactor plant generations are considered. By present 8 nuclear ice-breakers and a nuclear-powered cargo ship (lighter carrier) have been constructed, for which three of propulsion NSSS generations were developed. Their brief description, main performance indicators and results of operation since 1959 are given. It is shown that gradual evolution of NSSS design features has ensured creation of reliable, safe and environmentally friendly propulsion reactor plants. Issues of the propulsion NSSS life extension and improvement for new generation of nuclear ice-breakers, cargo ships, floating heat and power plants, sea water desalination and power generating complexes are considered with account for the gained operating experience. Activities on creation of new generation nuclear-powered ships and floating NPPs are considered as prospective sphere of Russia's collaboration with other countries of the World community. (author)

Higher plant availability and reduced reactor scram frequency in PWRs by appropriate system and I and C design

International Nuclear Information System (INIS)

Frei, G.; Weber, J.

1987-01-01

High plant availability and reliability are guaranteed by appropriate design of reactor and BOP systems, this including the plant I and C systems. It is of advantage to have design, construction and commissioning of the plant concentrated in the hands of a single company to avoid interface problems between the different areas of the plant. The integrated overall control concept developed by KWU with control, limitation and protection systems as well as optimized operational and monitoring systems assisted by instrumentation channel redundance and logic for selection of the second highest (or second lowest) signal value as appropriate for comparison with limitation setpoints, minimize the severity of transients. This results in a reduction in the frequency of reactor scrams and of unnecessary actuation of safety systems. Dynamic plant behavior is described for a number of examples where the improved plant behavior resulting from the above design features enhances plant availability

Process planning and accuracy distribution of marine power plant modularization

Directory of Open Access Journals (Sweden)

ZHANG Jinguo

2018-02-01

Full Text Available [Objectives] Modular shipbuilding can shorten the cycle of design and construction, lower production costs and improve the quality of products, but higher shipbuilding capabilities are required, especially for the installation of power plants. Because of such characteristics of modular shipbuilding as the high precision of docking links, long size equipment installation chain and quantitative docking interfaces, docking installation is very difficult due to high docking deviation and low accuracy of docking installation, leading to the abnormal vibration of equipment. In order to solve this problem, [Methods] on the basis of domestic shipbuilding capability, numerical calculation methods are used to analyze the accuracy distribution of modular installation. [Results] The results show that the accuracy distribution of different docking links is reasonable and feasible, and the setting of adjusting allowance matches the requirements of shipbuilding. [Conclusions] This method provides a reference for the modular construction of marine power plants.

Plant life management processes and practices for heavy water reactors

International Nuclear Information System (INIS)

Kang, K.-S.; Cleveland, J.; Clark, C.R.

2006-01-01

In general, heavy water reactor (HWR) nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. Their decisions are depending on essentially business model. They involve the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations. Continued plant operation, including operation beyond design life, called 'long term operation, depends, among other things, on the material condition of the plant. This is influenced significantly by the effectiveness of ageing management. Key attributes of an effective plant life management program include a focus on important systems, structure and components (SSCs) which are susceptible to ageing degradation, a balance of proactive and reactive ageing management programmes, and a team approach that ensures the co-ordination of and communication between all relevant nuclear power plant and external programmes. Most HWR NPP owners/operators use a mix of maintenance, surveillance and inspection (MSI) programs as the primary means of managing ageing. Often these programs are experienced-based and/or time-based and may not be optimised for detecting and/or managing ageing effects. From time-to-time, operational history has shown that this practice can be too reactive, as it leads to dealing with ageing effects (degradation of SSCs) after they have been detected. In many cases premature and/or undetected ageing cannot be traced back to one specific reason or an explicit error. The root cause is often a lack of communication, documentation and/or co-ordination between design, commissioning, operation or maintenance organizations. This lack of effective communication and interfacing frequently arises because, with the exception of major SSCs, such as the fuel channels or steam generators, there is a lack of explicit

Water treatment for the ISER [intrinsically safe and economical reactor] plant

International Nuclear Information System (INIS)

Sugawara, Ichiro.

1985-01-01

The ISER reactor assures inherent safety by causing the core, which is submerged in pool water containing a high boric acid concentration, to quickly shut down the nuclear reaction when overheating, pump trip or other problems occur. However, large quantities of pool water may cause difficulties in water quality control and waste management, resulting in higher costs. Therefore, the ISER as a total plant would not be publicly acceptable unless the water treatment and waste management system offer both safety balanced with reactor inherent safety, and economy counterbalanced by large quantities of pool water. This report clarifies the passive safety concept of possible waste treatment and management systems, and the ways to economically construct such facilities

Mirror fusion reactors

International Nuclear Information System (INIS)

Carlson, G.A.; Moir, R.W.

1978-01-01

We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

Mirror fusion reactors

International Nuclear Information System (INIS)

Anon.

1978-01-01

Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

Importance of benthonic marine flora monitoring in the liquid effluent discharge form Angra-1 Nuclear Power Plant

International Nuclear Information System (INIS)

Bloise, G.C.; Araujo Costa, D. de

1994-01-01

Angra-1 Nuclear Power Plant use sea water to condenser the steam of the secondary circuit. This water capted from Itaorna bay is chlorined and discharged more heater in Piraquara de Fora small bay. The temperature, chlorinade concentration, marine flora and fauna are monitored frequently with the intend of value the impact caused by this discharge to marine environment. The macroscopic marines algae is very sensible to environment temperature variations, constitutes on of the main rink in the food chain and stay every time attach at the bottom. Because of this facts they are considered an important bio indicators. (author). 5 refs, 2 figs, 2 tabs

Criteria for selecting measures of plant information with application to nuclear reactors

International Nuclear Information System (INIS)

Kiguchi, T.; Sheridan, T.B.

1979-01-01

Criteria for selecting from alternative measures of plant information have been derived by information theory and the expected-value model of decision theory. Two criteria derived by information theory are minimum ''uncertainty'' of judgement of plant states, and maximum ''surprise'' of human operators. Two criteria derived by decision theory are maximum expected reward to operators, and maximum ratio of the maximum and next-to-maximum expected rewards. These four criteria have been evaluated experimentally in application to human monitoring problems of three ''plants'': an abstract laboratory exercise, a small research reactor, and a conventional nuclear power plant (high-fidelity training simulator). The results show that the criteria derived by decision theory are superior to the others from the viewpoint of correct rate of human subjects' judgement

Assessment and management of ageing of major nuclear power plant components important to safety: CANDU reactor assemblies

International Nuclear Information System (INIS)

2001-02-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring, and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs) including the Soviet designed water moderated and water cooled energy reactors (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which

Considerations on safety against seismic excitations in the project of reactor auxiliary building and control building in nuclear power plants

International Nuclear Information System (INIS)

Santos, S.H.C.; Castro Monteiro, I. de

1986-01-01

The seismic requests to be considered in the project of main buildings of a nuclear power plant are discussed. The models for global seismic analysis of nuclear power plant structures, as well as models for global strength distribution are presented. The models for analysing reactor auxiliary building and control building, which together with the reactor building and turbine building form the main energy generation complex in a nuclear power plant, are described. (M.C.K.) [pt

Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

2013-10-15

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

Directory of Open Access Journals (Sweden)

VIKAS JAIN

2013-10-01

Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

Knowledges and abilities catalog for nuclear power plant operators: pressurized water reactors

International Nuclear Information System (INIS)

1985-07-01

This document catalogs roughly 5300 knowledges and abilities of reactor operators and senior reactor operators. It results from a reanalysis of much larger job-task analysis data base compiled by the Institute of Nuclear Power Operations (INPO). Knowledges and abilities are cataloged for 45 major power plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functions (e.g., reactivity control and reactor coolant system inventory control). With appropriate sampling from this catalog, operator licensing examinations having content validity can be developed. A structured sampling procedure for this catalog is under development by the Nuclear Regulatory Commission (NRC) and will be published as a companion document, ''Examiners' Handbook for Developing Operator Licensing Examinations'' (NUREG-1121). The examinations developed by using the catalog and handbook will cover those topics listed under Title 10, Code of Federal Regulations, Part 55

Dynamic analysis of the condensate feedwater system in boiling water reactor plants

International Nuclear Information System (INIS)

Tanji, J.; Omori, T.

1982-01-01

The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

Reactor type choice and characteristics for a small nuclear heat and electricity co-generation plant

International Nuclear Information System (INIS)

Liu Kukui; Li Manchang; Tang Chuanbao

1997-01-01

In China heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NO x , SO x and CO 2 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection. The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application. The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down. In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents. (author)

Comparative distributions of alkaline earths and Pb among tissues of marine plants and animals

International Nuclear Information System (INIS)

Burnett, M.W.; Settle, D.M.; Patterson, C.C.

1978-01-01

Lead, barium, strontium and calcium were studied by isotope dilution, clean-lab techniques in both a marine and a terrestrial ecosystem. Analyses for Pb and Ba are difficult since their concentrations range down to the ng g -1 level in plant and animal tissue. Experimental details are given. Results are presented and discussed. (U.K.)

Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

Science.gov (United States)

Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

2017-03-01

The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

Economic competitiveness of small modular reactors versus coal and combined cycle plants

International Nuclear Information System (INIS)

Alonso, Gustavo; Bilbao, Sama; Valle, Edmundo del

2016-01-01

Small modular reactors (SMRs) may be an option to cover the electricity needs of isolated regions, distributed generation grids and countries with small electrical grids. Previous analyses show that the overnight capital cost for SMRs is between 4500 US$/kW and 5350 US$/kW, which is between a 6% and a 26% higher than the average cost of a current large nuclear reactor. This study analyzes the economic competitiveness of small modular reactors against thermal plants using coal and natural gas combined cycle plants. To assess the economic competitiveness of SMRs, three overnight capital costs are considered 4500 US$/kW, 5000 US$/kW and 5350 US$/kW along with three discount rates for each overnight cost considered, these are 3, 7, and 10%. To compare with natural gas combined cycle (CC) units, four different gas prices are considered, these are 4.74 US$/GJ (5 US$/mmBTU), 9.48 US$/GJ (10 US$/mmBTU), 14.22 US$/GJ (15 US$/mmBTU), and 18.96 US$/GJ (20 US$/mmBTU). To compare against coal, two different coal prices are considered 80 and 120 US$/ton of coal. The carbon tax considered, for both CC and coal, is 30 US$/ton CO_2. The results show what scenarios make SMRs competitive against coal and/or combined cycle plants. In addition, because the price of electricity is a key component to guarantee the feasibility of a new project, this analysis calculates the price of electricity for the economically viable deployment of SMRs in all the above scenarios. In particular, this study shows that a minimum price of electricity of 175 US$/MWh is needed to guarantee the feasibility of a new SMR, if its overnight capital cost is 5350 US$/kWe and the discount rate is 10%. Another result is that when the price of electricity is around 100 US$/MWh then the discount rate must be around 7% or less to provide appropriate financial conditions to make SMRs economically feasible. - Highlights: • Small modular reactor (SMR) are economically assessed. • SMR are compared against gas and coal

The obscure factor analysis on the vibration reliability of the internals of nuclear power plant reactor and anti-vibration measures

International Nuclear Information System (INIS)

Fu Geyan; Zhu Qirong

1998-11-01

It is pointed out that the main reason making nuclear power plants reactors leak is the vibration of internals of reactors. The factors which lead the vibration all have randomness and obscureness. The obscure reliability theory is introduced to the vibration system of internals of nuclear power reactor. Based on a quantity of designing and moving data, the obscure factors effecting the vibration reliability of the internals of nuclear power plant reactor are analyzed and the anti-vibration reliability criteria and the evaluating model are given. And the anti-vibration reliability measures are advanced from different quarters of the machine design and building, the thermohydraulics design, the control of reactivity, etc.. They may benefit the theory and practice for building and perfecting the vibration obscure reliability model of the reactor internals

Temperature conditions of foundation plates under nuclear power plant reactor compartments

International Nuclear Information System (INIS)

Ehsaulov, S.L.

1990-01-01

Method for calculation of temperature conditions for foundation plates under reactor compartments located in the main building, used in construction of the second stage of the Kostroma nuclear power plant, is considered. The obtained calculation data can be used for determining the most suitable period of concrete placement, composition, initial temperature, manufacturing technology and ways of delivery of concrete mixture

Pump selection and application in a pressurized water reactor electric generating plant

International Nuclear Information System (INIS)

Kitch, D.M.

1985-01-01

Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

Repair and replacement of reactor internals for plant life extension

International Nuclear Information System (INIS)

Graae, T.

1998-01-01

Recent experience from early Swedish BWRs corroborate that all components in a nuclear power plant can be repaired or replaced with new ones. Oskarshamn 1 has gone through a thorough refurbishment project. A number of internals were repaired or replaced including the core shroud support which was welded to the bottom of the reactor pressure vessel. The project verifies that it is fully possible to carry out complicated inspection and repair work inside a nuclear pressure vessel which has been in operation for more than 20 years. Along with increased capacity factor, operating nuclear power plants get the financial conditions needed for extensive repair and modernization projects. Large power output leads to short pay-back times for the investments. The FENIX project at Oskarshamn 1 is such a project. There are utilities whose policy is to keep their plants in as-new condition for an unlimited length of time. (orig.)

Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Reactor safety assessment system

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

FASTER Test Reactor Preconceptual Design Report

Energy Technology Data Exchange (ETDEWEB)

Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-31

The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

International Nuclear Information System (INIS)

Galvin, Mark R.; Todreas, Neil E.; Conway, Larry E.

2003-01-01

New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented

Normal and compact spent fuel storage in light water reactor power plants

International Nuclear Information System (INIS)

Kuenel, R.R.

1978-01-01

The compact storage of light water reactor spent fuel is a safe, cheap and reliable contribution towards overcoming the momentarily existing shortage in spent fuel reprocessing. The technical concept is described and physical behaviour discussed. The introduction of compact storage racks in nuclear power plants increases the capacity from 100 to about 240 %. The increase in decay heat is not more than about 14%, the increase in activity inventory and hazard potential does not exceed 20%. In most cases the existing power plant equipment fulfils the new requirements. (author)

Fracture toughness requirements of reactor vessel material in evaluation of the safety analysis report of nuclear power plants

International Nuclear Information System (INIS)

Widia Lastana Istanto

2011-01-01

Fracture toughness requirements of reactor vessel material that must be met by applicants for nuclear power plants construction permit has been investigated in this paper. The fracture toughness should be described in the Safety Analysis Reports (SARs) document that will be evaluated by the Nuclear Energy Regulatory Agency (BAPETEN). Because BAPETEN does not have a regulations or standards/codes regarding the material used for the reactor vessel, especially in the fracture toughness requirements, then the acceptance criteria that applied to evaluate the fracture toughness of reactor vessel material refers to the regulations/provisions from the countries that have been experienced in the operation of nuclear power plants, such as from the United States, Japan and Korea. Regulations and standards used are 10 CFR Part 50, ASME and ASTM. Fracture toughness of reactor vessel materials are evaluated to ensure compliance of the requirements and provisions of the Regulatory Body and the applicable standards, such as ASME or ASTM, in order to assure a reliability and integrity of the reactor vessels as well as providing an adequate safety margin during the operation, testing, maintenance, and postulated accident conditions over the reactor vessel lifetime. (author)

PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

International Nuclear Information System (INIS)

Vondy, D.R.

1979-10-01

The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined

PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1979-10-01

The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

Energy Technology Data Exchange (ETDEWEB)

Fang, Maoyao; Peng, Minjun; Cheng Shouyu [Harbin Engineering University, Harbin (China)

2014-08-15

In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions.

The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

International Nuclear Information System (INIS)

Fang, Maoyao; Peng, Minjun; Cheng Shouyu

2014-01-01

In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions

Development of graphic display program of reactor operating parameters for emergency exercise at nuclear power plants

International Nuclear Information System (INIS)

Okuda, Yasunori; Yoshida, Yoshitaka; Gotou, Kazuko

2001-01-01

A scenario of nuclear emergency exercise based on the result of accident progress analysis is expected to ensure effective training. Thereupon a new graphic display program for reactor operating parameters has been developed to present real-time of plant process values (parameters), released radioactivities from the plant, and dose rate data around the site calculated by using the accident analysis code MAAP4 and other codes. This system has a trend graph screen displaying reactor operating parameters, an environmental dose rate summary screen indicating dose rate distribution around the site on the map, and a plant parameters summary screen showing important plant parameters on a simplified plant system diagram. One screen can be switched to another any time. It also has a jump-function easily accessing any stage during the exercise scenario in accordance with progress of the exercise. As a result of the application of this system to a real nuclear emergency exercise, it has been verified that this system is quite useful for confirming the parameters when the nuclear emergency exercise starts and the licensee reports the plant conditions to related bodied. (author)

Development of graphic display program of reactor operating parameters for emergency exercise at nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Okuda, Yasunori; Yoshida, Yoshitaka [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan); Gotou, Kazuko [Kansai Electric Power Co., Inc., Osaka (Japan)

2001-09-01

A scenario of nuclear emergency exercise based on the result of accident progress analysis is expected to ensure effective training. Thereupon a new graphic display program for reactor operating parameters has been developed to present real-time of plant process values (parameters), released radioactivities from the plant, and dose rate data around the site calculated by using the accident analysis code MAAP4 and other codes. This system has a trend graph screen displaying reactor operating parameters, an environmental dose rate summary screen indicating dose rate distribution around the site on the map, and a plant parameters summary screen showing important plant parameters on a simplified plant system diagram. One screen can be switched to another any time. It also has a jump-function easily accessing any stage during the exercise scenario in accordance with progress of the exercise. As a result of the application of this system to a real nuclear emergency exercise, it has been verified that this system is quite useful for confirming the parameters when the nuclear emergency exercise starts and the licensee reports the plant conditions to related bodied. (author)

Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

International Nuclear Information System (INIS)

Toft, P.; Borring, J.; Adolph, E.

1988-01-01

A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

Catalogue and classification of technical safety rules for light-water reactors and reprocessing plants

International Nuclear Information System (INIS)

Bloser, M.; Fichtner, N.; Neider, R.

1975-08-01

This report on the cataloguing and classification of technical rules for land-based light-water reactors and reprocessing plants contains a list of classified rules. The reasons for the classification system used are given and discussed

Sono-photo-degradation of carbamazepine in a thin falling film reactor: Operation costs in pilot plant.

Science.gov (United States)

Expósito, A J; Patterson, D A; Monteagudo, J M; Durán, A

2017-01-01

The photo-Fenton degradation of carbamazepine (CBZ) assisted with ultrasound radiation (US/UV/H 2 O 2 /Fe) was tested in a lab thin film reactor allowing high TOC removals (89% in 35min). The synergism between the UV process and the sonolytic one was quantified as 55.2%. To test the applicability of this reactor for industrial purposes, the sono-photo-degradation of CBZ was also tested in a thin film pilot plant reactor and compared with a 28L UV-C conventional pilot plant and with a solar Collector Parabolic Compound (CPC). At a pilot plant scale, a US/UV/H 2 O 2 /Fe process reaching 60% of mineralization would cost 2.1 and 3.8€/m 3 for the conventional and thin film plant respectively. The use of ultrasound (US) produces an extra generation of hydroxyl radicals, thus increasing the mineralization rate. In the solar process, electric consumption accounts for a maximum of 33% of total costs. Thus, for a TOC removal of 80%, the cost of this treatment is about 1.36€/m 3 . However, the efficiency of the solar installation decreases in cloudy days and cannot be used during night, so that a limited flow rate can be treated. Copyright © 2016 Elsevier B.V. All rights reserved.

Development of analytical code `ACCORD` for incore and plant dynamics of High Temperature Gas-cooled Reactor

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Itakura, Hirofumi

1996-11-01

Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The `ACCORD` code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the `ACCORD` code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)

Development of analytical code 'ACCORD' for incore and plant dynamics of High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko; Itakura, Hirofumi.

1996-11-01

Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The 'ACCORD' code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the 'ACCORD' code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)

Development of digital plant protection system for Korean Next Generation Reactor

International Nuclear Information System (INIS)

Suk-Joon Park

1998-01-01

A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author)

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

Integrated plant safety assessment: Systematic Evaluation Program. LaCrosse Boiling Water Reactor, Dairyland Power Cooperative, Docket No. 50-409

International Nuclear Information System (INIS)

1983-04-01

The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addresed. Equipment and procedural changes have been identified as a result of the review

Reactor safety study applied to the Forsmark 3 Power Plant

International Nuclear Information System (INIS)

Ericsson, G.; Tiren, L.I.

1978-01-01

A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

Transportable nuclear power plant T3C-M with two reactor plants of improved safety

International Nuclear Information System (INIS)

Ogloblin, B.G.; Gromov, B.F.

1993-01-01

Development and cultivating of districts in Siberia, North, Far East, Kamchatka and other remote or almost inaccessible district of the country depends to a large degree on their providing with power. The specific character of these districts imposes in turn a wide variety of special requirements upon the power sources. In particular, it is essential to provide the following; maximum manufacture availability of the whole equipment at the minimum volume of construction and installation work on operation site, high safety, longterm service life, ecologically, minimum scope of work on equipment in-service maintenance and inspection, etc. Taking into account the well-known difficulties connected with the delivery of conventional energy carriers to the above-mentioned districts and the situation with the alternative power sources, the application of the low-power nuclear plants (NPP) for these purposes looks definitely promising. Among the probable trends in creating the NPPs of this type as very promising is considered the possibility to apply the two-circuit reactor plant of the vessel type with the liquid lead as a primary coolant and free air as a secondary coolant and working medium in the open gas-turbine cycle. The nuclear plant T3C-M of improved safety with two of this type reactor plants with total electric power of 8 MW is developed by CDB of Machine Building with participation of several enterprises of St. Petersburg under the scientific leadership and is intended for generation of electric power and up to 4 Gcal/h of heat for populated areas and installations placed at long distance from the main electric power supply sources where it is difficult or non-efficient economically to deliver the conventional kinds of fuel. The main principles being laid as a basis when developing the proposed NPP will allow one to create mobile power sources which possess a high degree of safety and inherent self-protection

Demonstration of a 100-kWth high-temperature solar thermochemical reactor pilot plant for ZnO dissociation

Science.gov (United States)

Koepf, E.; Villasmil, W.; Meier, A.

2016-05-01

Solar thermochemical H2O and CO2 splitting is a viable pathway towards sustainable and large-scale production of synthetic fuels. A reactor pilot plant for the solar-driven thermal dissociation of ZnO into metallic Zn has been successfully developed at the Paul Scherrer Institute (PSI). Promising experimental results from the 100-kWth ZnO pilot plant were obtained in 2014 during two prolonged experimental campaigns in a high flux solar simulator at PSI and a 1-MW solar furnace in Odeillo, France. Between March and June the pilot plant was mounted in the solar simulator and in-situ flow-visualization experiments were conducted in order to prevent particle-laden fluid flows near the window from attenuating transparency by blocking incoming radiation. Window flow patterns were successfully characterized, and it was demonstrated that particle transport could be controlled and suppressed completely. These results enabled the successful operation of the reactor between August and October when on-sun experiments were conducted in the solar furnace in order to demonstrate the pilot plant technology and characterize its performance. The reactor was operated for over 97 hours at temperatures as high as 2064 K; over 28 kg of ZnO was dissociated at reaction rates as high as 28 g/min.

Investigation of small and modular-sized fast reactor

International Nuclear Information System (INIS)

Kubota, Kenichi; Kawasaki, Nobuchika; Umetsu, Yoichiro; Akatsu, Minoru; Kasai, Shigeo; Konomura, Mamoru; Ichimiya, Masakazu

2000-06-01

In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

Initiating Events for Multi-Reactor Plant Sites

Energy Technology Data Exchange (ETDEWEB)

Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-09-01

Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

International Nuclear Information System (INIS)

Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

2017-01-01

This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

Transuranic elements in marine environments

International Nuclear Information System (INIS)

Bowen, V.T.

1975-04-01

Transuranic elements are present in marine environments as a result of worldwide fallout, close-in fallout, the SNAP-9A burnup, pipeline disposal of reprocessing wastes, neutron capture by uranium in one-pass cooling-water reactors and the B-52 crash in Thule, Greenland. Distributions and movements of 239 Pu, 240 Pu, 241 Am, and 137 Cs from these introductions in the Atlantic Ocean are being studied partly because the transuranic elements themselves are geochemically interesting, partly because they appear to serve as tracers for specific oceanographic processes, and partly because of concern that man is faced with the problem of disposal of rapidly increasing amounts of transuranics as radioactive waste, and that we must be able, soon, to predict the fates and the effects of those amounts that reach the coastal waters or the deep oceans. Plutonium and americium are widely distributed in the oceans as a result of man's activities. Both appear to be more mobile than expected, and Pu shows little behavior in these environments that had been predicted from laboratory studies. Although their associations with biological material seem to be most striking for rooted plants or Sargassum, it is too premature to dismiss the possibility of their being a real hazard to marine life

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

UF6 breeder reactor power plants for electric power generation

International Nuclear Information System (INIS)

Rust, J.H.; Clement, J.D.; Hohl, F.

1976-01-01

The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

Extended fuel cycle operation for pressurized water reactor plants

International Nuclear Information System (INIS)

Silvestri, G.J. Jr.

1978-01-01

A nuclear steam turbine power plant system having an arrangement therein for extended fuel cycle operation is described. The power plant includes a turbine connected at its inlet to a source of motive fluid having a predetermined pressure associated therewith. The turbine has also connected thereto an extraction conduit which extracts steam from a predetermined location therein for use in an associated apparatus. A bypass conduit is provided between a point upstream of the inlet and the extraction conduit. A flow control device is provided within the bypass conduit and opens when the pressure of the motive steam supply drops beneath the predetermined pressure as a result of reactivity loss within the nuclear reactor. Opening of the bypass conduit provides flow to the associated apparatus and at the same time provides an increased flow orifice to maintain fluid flow rate at a predetermined level

Various mortars for anti-fouling purposes in marine environments

Energy Technology Data Exchange (ETDEWEB)

Kanematsu, Hideyuki; Masuda, Tomoka [Department of Materials Science and Engineering, Suzuka National College of Technology, Shiroko-cho, Suzuka, Mie 510-0294 (Japan); Miura, Yoko; Kuroda, Daisuke [Department of General Education, The Company, Suzuka National College of Technology, Shiroko-cho, Suzuka, Mie 510-0294 (Japan); Hirai, Nobumitsu [Department of Chemistry and Biochemistry, Suzuka National College of Technology, Shiroko-cho, Suzuka, Mie 510-0294 (Japan); Yokoyama, Seiji [Department of Mechanical Engineering, Toyohashi University of Technology, 1-1, Hibarigaoka, Tenpaku, Toyohashi, Aichi, 441-8580 (Japan)

2014-02-20

The antifouling properties for some mortars with steel making slags were investigated by real marine immersion tests and a unique laboratory acceleration tests with a specially devised biofilm acceleration reactors. Mortars mixed with steel making slags containing abundant iron elements tended to form biofilm and also bifouling. The two kinds of biofilm formation tests were used in this study. Real immersion in marine environments and laboratory test with a specially devised biofilm acceleration reactor. The former evaluated the biofouling characteristics more properly, while the latter did the biofilm formation characteristics more effectively.

Various mortars for anti-fouling purposes in marine environments

International Nuclear Information System (INIS)

Kanematsu, Hideyuki; Masuda, Tomoka; Miura, Yoko; Kuroda, Daisuke; Hirai, Nobumitsu; Yokoyama, Seiji

2014-01-01

The antifouling properties for some mortars with steel making slags were investigated by real marine immersion tests and a unique laboratory acceleration tests with a specially devised biofilm acceleration reactors. Mortars mixed with steel making slags containing abundant iron elements tended to form biofilm and also bifouling. The two kinds of biofilm formation tests were used in this study. Real immersion in marine environments and laboratory test with a specially devised biofilm acceleration reactor. The former evaluated the biofouling characteristics more properly, while the latter did the biofilm formation characteristics more effectively

Training simulator for nuclear power plant reactor control model and method

International Nuclear Information System (INIS)

Czerbuejewski, F.R.

1975-01-01

A description is given of a method and system for the real-time dynamic simulation of a nuclear power plant for training purposes, wherein a control console has a plurality of manual and automatic remote control devices for operating simulated control rods and has indicating devices for monitoring the physical operation of a simulated reactor. Digital computer means are connected to the control console to calculate data values for operating the monitoring devices in accordance with the control devices. The simulation of the reactor control rod mechanism is disclosed whereby the digital computer means operates the rod position monitoring devices in a real-time that is a fraction of the computer time steps and simulates the quick response of a control rod remote control lever together with the delayed response upon a change of direction

Nuclear power plant design characteristics. Structure of nuclear power plant design characteristics in the IAEA Power Reactor Information System (PRIS)

International Nuclear Information System (INIS)

2007-03-01

One of the IAEA's priorities has been to maintain the Power Reactor Information System (PRIS) database as a viable and useful source of information on nuclear reactors worldwide. To satisfy the needs of PRIS users as much as possible, the PRIS database has included also a set of nuclear power plant (NPP) design characteristics. Accordingly, the PRIS Technical Meeting, organized in Vienna 4-7 October 2004, initiated a thorough revision of the design data area of the PRIS database to establish the actual status of the data and make improvements. The revision first concentrated on a detailed review of the design data completion and the composition of the design characteristics. Based on the results of the review, a modified set and structure of the unit design characteristics for the PRIS database has been developed. The main objective of the development has been to cover all significant plant systems adequately and provide an even more comprehensive overview of NPP unit designs stored in the PRIS database

Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Executive summary: main report

International Nuclear Information System (INIS)

1975-10-01

Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks

Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendices VII, VIII, IX, and X

International Nuclear Information System (INIS)

1975-10-01

Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy

Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

International Nuclear Information System (INIS)

Adam, E.; Carl, H.; Kubis, K.

1979-01-01

Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

Marine environment status assessment based on macrophytobenthic plants as bio-indicators of heavy metals pollution.

Science.gov (United States)

Zalewska, Tamara; Danowska, Beata

2017-05-15

The main aim of study was to develop the environmental quality standards (EQS MP ) for selected heavy metals: Pb, Cd, Hg and Ni bioaccumulated in the tissues of marine macrophytobenthic plants: Chara baltica, Cladophora spp., Coccotylus truncatus, Furcellaria lumbricalis, Polysiphonia fucoides, Stuckenia pectinata and Zanichellia palustris, collected in designated areas of the southern Baltic Sea in period 2008-2015. The calculated concentration ratios (CR), which attained very high values: 10 4 Lkg -1 for lead, 10 3 Lkg -1 for nickel and mercury and even 10 5 Lkg -1 for cadmium formed the basis for the determination of EQS MP values. The EQS MP values were: 26mgkg -1 d.w. for Pb, 33mgkg -1 d.w. for Cd, 32mgkg -1 d.w. for Ni and 0.4mgkg -1 d.w. for Hg. The application of macrophytobenthic plants as bioindicators in marine environment status assessment of certain areas of the Baltic Sea is also described in the paper. Copyright © 2017 Elsevier Ltd. All rights reserved.

FASTER test reactor preconceptual design report summary

Energy Technology Data Exchange (ETDEWEB)

Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-02-29

The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

Seashore marine table quiz

OpenAIRE

Institute, Marine

2013-01-01

Develop an increasing awareness of plants and animals that live in local marine environments including the seashore, seas and oceans of Ireland. After learning all about the seashore and other marine related lessons, this quiz can be used to evaluate the student’s knowledge of the marine related living things and natural environments. The table quiz can be used as a guide, highlighting facts about the marine environment and some of the animals that live there.

On the reliability of steam generator performance at nuclear power plants with WWER type reactors

International Nuclear Information System (INIS)

Styrikovich, M.A.; Margulova, T.Kh.

1974-01-01

The problem of ensuring reliable operation of steam generators in a nuclear power plant with a water-cooled, water-moderated reactor (WWER) was studied. At a nuclear power plant with a vertical steam generator (specifically, a Westinghouse product) the steam generator tubes were found to have been penetrated. Shutdown was due to corrosion disintegration of the austenitic stainless steel, type 18/8, used as pipe material for the heater surface. The corrosion was the result of the action of chlorine ions concentrated in the moisture contained in the iron oxide films deposited in low parts of the tube bundle, directly at the tube plate. Blowing through did not ensure complete removal of the film, and in some cases the construction features of the steam generator made removal of the film practically impossible. Replacement of type 18/8 stainless steel by other construction material, e.g., Inconel, did not give good results. To ensure reliable operation of vertical steam generators in domestic practice, the generators are designed without a low tube plate (a variant diagram of the vertical steam generator of such construction for the water-cooled, water-moderated reactor 1000 is presented). When low tube plates are used the film deposition is intolerable. For organization of a non-film regime a complex treatment of the feed water is used, in which the amount of complexion is calculated from the stoichmetric ratios with the composition of the feed water. It is noted that, if 100% condensate purification is used with complexon processing of the feed water to the generator, we can calculate the surface of the steam-generator heater without considering the outer placement on the tubes. In this the cost of the steam generator and all the nuclear power plants with WWER type reactors is decreased even with installation of a 100% condensate purification. It is concluded that only simultaneous solution of construction and water-regime problems will ensure relaible operation of

Analysis of chemical factors affecting marine ecosystem around nuclear power plant

International Nuclear Information System (INIS)

Chun, Kwan Sik; Choi, Yoon Dong; Chun, Ki Jeong; Kim, Jin Kyu; Jung, Kyeong Chai; Lee, Yeong Keun; Park, Hyo Kook

1994-06-01

The ecological data of the coastal area of Youngkwang nuclear power plant from 1987 to 1993 were comprehensively analyzed, and various physical and chemical properties of sea water and sediments were measured. Major factors affecting phytoplankton standing crops were suspended substances, nitrate, and silicate. The contents of iron, chromium, copper, and sulfur in sediments sampled from the discharge channel were slightly higher than those in the other areas. In order to qantify the chemical impacts on marine ecosystem, it is desirable that a systematic survey be made through the whole year cycle to assure the consistency and confidence of the related data. (Author)

Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

Science.gov (United States)

Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

2016-01-01

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

System Evaluation and Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen-Production Plant

International Nuclear Information System (INIS)

Harvego, E.A.; McKellar, M.G.; Sohal, M.S.; O'Brien, J.E.; Herring, J.S.

2010-01-01

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current (AC) to direct current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

Development of digital plant protection system for Korean Next Generation Reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Suk-Joon [NSSS Engineering and Development Division, Korea Power Engineering Company, Taejon (Korea, Republic of)

1998-10-01

A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author) 8 refs, 4 figs, 3 tabs

Civilian application of propulsion reactor in Indonesia

International Nuclear Information System (INIS)

Djokolelono, M.; Arbie, B.; Lasman, A.N.

2000-01-01

It has been learned that to cope with energy requirement in the remote islands and less developed regions of Indonesia, small or very small nuclear reactors producing electricity and/or process heat could be appropriately applied. The barge mounted propulsion power reactors are the attractive examples so far envisioned, and technology information of which is being exposed to the world these last years. The solutions for least maintenance and no on-site refueling, no radioactive discharge, and no radioactive waste to remain in the user country are among the attractions for further deliberations. It has been understood, however, that there are many uncertainties to overcome, especially for the developing countries to introduce this novel application. International acceptance is the most crucial, availability of first-of-the-kind engineering, prototype or reference plant that would prove licensibility in the vendor's country is the second, and economic competitiveness due to very small size is the third among issues to enlighten. The relevant regulations concerning marine nuclear safety, marine transportation, and proliferation of information, as well as international forums to justify the feasibility of related transfer of technology, are the items that the IAEA could help to provide to smoothen any possible international transaction. Indonesia supports this AGM as one of the appropriate IAEA efforts in this line, and expects from it positive international consensus and possible studies/R and D work that this country could participate in. (author)

Realization of safety culture into a reactor plant-4S (super safe, small and simple) LMR

International Nuclear Information System (INIS)

Hattori, S.; Ikemoto, I.; Minato, A.

1998-01-01

International Nuclear Safety Advisory Group (INSAG) defines Safety Culture as the following; Safety Culture is that assembly of characteristics and attitudes in organization and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance. While such Safety Culture is certainly a critical element of nuclear safety assurance, it is important to design nuclear power facilities as friendly to operators as possible with minimum dependence on human factors. From the viewpoint of ensuring supply in our global society, it will be necessary to have multiple approaches to further promote the use of nuclear energy worldwide despite various social and cultural restrictions. It should then be considered, as prospective options, to disperse small nuclear power plants throughout the world under technical, social and cultural conditions. Under this circumstance, we have quested for and now propose a a scheme of assuring sheer safety of nuclear power plants by implementing operator-friendly nuclear reactors virtually free from human errors. The scheme specifically includes the measures for improving reliability through fabricating more compact reactors with a continuous in-factory production system, simplify maintenance and inspection of the reactor system using passive systems and further relieving operators of burden of labors. (author)

The materials concept in German light water reactors. A contribution to plant safety, economic performance and damage prevention

International Nuclear Information System (INIS)

Ilg, Ulf

2008-01-01

Major decisions taken as early as in the planning and construction phases of nuclear power plants may influence overall plant life. Component quality at the beginning of plant life is determined very much also by a balanced inclusion of the 'design, choice of materials, manufacturing and inspection' elements. One example of the holistic treatment of design, choice of material, and manufacture of important safety-related components in pressurized water reactors is the reactor pressure vessel (RPV) in which the ferritic compound tubes, with inside claddings, for the control rod drive nozzles are screwed into the vessel top. Also the choice of Incoloy 800 for the steam generator tubes, and the design of the main coolant pipes with inside claddings as seamless pipe bends / straight pipes with integrated nozzles connected to mixed welds with austenitic pipes are other special design features of the Siemens/KWU plants. A demonstrably high quality standard by international comparison to this day has been exhibited by the austenitic RPV internals of boiling water reactors, which were made of a low-carbon Nb-stabilized austenitic steel grade by optimum manufacturing technologies. The same material is used for backfitting austenitic pipes. Reliable and safe operation of German nuclear power plants has been demonstrated for more than 4 decades. One major element in this performance is the materials concept adopted in Germany also in the interest of damage prevention. (orig.)

Structural analysis of the Upper Internals Structure for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Houtman, J.L.

1979-01-01

The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides control of core outlet flow to prevent severe thermal transients from occuring at the reactor vessel and primary heat transport outlet piping, provides instrumentation to monitor core performance, provides support for the control rod drivelines, and provides secondary holddown of the core. All of the structural analysis aspects of assuring the UIS is structurally adequate are presented including simplified and rigorous inelastic analysis methods, elevated temperature criteria, environmental effects on material properties, design techniques, and manufacturing constraints

Development of RETRAN-03/MOV code for thermal-hydraulic analysis of nuclear reactor under moving conditions

International Nuclear Information System (INIS)

Kim, Hak Jae; Park, Goon Cherl

1996-01-01

Nuclear ship reactors have several; features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been performed under rolling,heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removed to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions. 7 refs., 11 figs. (author)

Fossil fuel furnace reactor

Science.gov (United States)

Parkinson, William J.

1987-01-01

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Applicability of base-isolation R ampersand D in non-reactor facilities to a nuclear reactor plant

International Nuclear Information System (INIS)

Seidensticker, R.W.; Chang, Y.W.

1990-01-01

Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R ampersand D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R ampersand D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R ampersand D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R ampersand D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

Medical surveillance of nuclear power plant workers during reactor shutdown using whole-body counting and excretion analysis

International Nuclear Information System (INIS)

Le Roux-Desmis, C.

1987-01-01

After a review of radioactivity basis and radiation protection principles, the various aspects of medical surveillance of nuclear power plant workers during reactor shutdown, are presented. Internal contamination incidents that happened during 1986-1987 shutdown of Paluel reactor are exposed. Internal contamination levels are evaluated using whole-body counting and radionuclide determination in feces and urine and compared with dose limits [fr

Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

International Nuclear Information System (INIS)

Woo, H.H.; Lu, S.C.

1981-01-01

Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

Research and engineering application of coordinated instrumentation control and protection technology between reactor and steam turbine generator on nuclear power plant

International Nuclear Information System (INIS)

Sun Xingdong

2014-01-01

The coordinated instrumentation control and protection technology between reactor and steam turbine generator (TG) usually is very significant and complicated for a new construction of nuclear power plant, because it carries the safety, economy and availability of nuclear power plant. Based on successful practice of a nuclear power plant, the experience on interface design and hardware architecture of coordinated instrumentation control and protection technology between reactor and steam turbine generator was abstracted and researched. In this paper, the key points and engineering experience were introduced to give the helpful instructions for the new project. (author)

An integral reactor design concept for a nuclear co-generation plant

International Nuclear Information System (INIS)

Lee, D.J.; Kim, J.I.; Kim, K.K.; Chang, M.H.; Moon, K.S.

1997-01-01

An integral reactor concept for nuclear cogeneration plant is being developed at KAERI as an attempt to expand the peaceful utilization of well established commercial nuclear technology, and related industrial infrastructure such as desalination technology in Korea. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway to evaluate the characteristics of various passive safety concepts and provide the proper technical data for the conceptual design. This paper describes the preliminary safety and design concepts of the advanced integral reactor. Salient features of the design are hexagonal core geometry, once-through helical steam generator, self-pressurizer, and seismic resistant fine control CEDMS, passive residual heat removal system, steam injector driven passive containment cooling system. (author)

Methane production from marine macroalgae. A literature study with comments

Energy Technology Data Exchange (ETDEWEB)

Josefsen, K.; Aasen, I.M.

1995-06-16

This report is a survey of the published literature on fermentation of marine macroalgae to produce methane. The emphasis is placed on modern fermentation process development, including both bio-technological and economic parameters. Marine macroalgae are mostly good feedstock for methane fermentation. The main carbohydrates in seaweeds are alginate, laminaria and mannitol. Both the N and P content of kelp are usually high enough to avoid nutrient limitation. The biogass produced from marine macroalgae usually contains 50 - 65% methane. Experimentally methane yields in the order of 0.35 - 0.43 m{sup 3}/kg volatile solids (VS) have been obtained from Macrocystis pyrifera and 0.20 - 0.30 m{sup 3}/kg VS for Laminaria sp. at long retention times (50-60 days) in completely mixed reactors. The maximum reported production rate in a completely mixed reactor is 2.7 vol CH{sub 4}/reactor volume x day for M. pyrifera. In reactor configurations giving longer solids retention times (SRT) than liquid retention times (LRT), the highest reported productivity was 3.3 vol/vol x day, with loading rate 9.6 kg VS/m{sup 3} x day, HRT 10 days and SRT 23 days. There are discussions of special problems related to fermentation of marine algae, in particular the sulphur content, the toxicity of H{sub 2}S and the precipitation of heavy metals as sulphides. 72 refs., 5 figs., 13 tabs.

A Reliability-Based Determination of Economic Life of Marine power plants

International Nuclear Information System (INIS)

Atua, K.

1999-01-01

The reliability-based life approach is utilized. Selective failure modes of marine power plants are used for illustration. A case study of the Egyptian Commercial Fleet owned by the Public Sector Company was analyzed and used to establish a demonstration of the expected economic life based on local operating and maintenance conditions. The data acquired is analyzed and failure trend is derived for each failure mode. Probabilistic techniques are used to randomly generate numbers and times of occurrence of different failure modes. The reliability analysis is performed on the life span expected by the manufacture to predict the total number of failures, dependent failures, and cost of failures. Total expenditure due to random failure and cost of scheduled maintenance together with the annual income are utilized (using the time value of money) to determine the economic life of the plant. Conclusions are derived and recommendations for the enhancement of this work in the future are made

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

Grover, S. Blaine

2009-01-01

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

Calculation of energetic characteristics of C-14 emitted from Beloyarsk nuclear power plant plume with fast neutron reactor

Science.gov (United States)

Kolotkov, Gennady A.; Penin, Sergei

2017-11-01

The paper examines an update of comparative analysis of radionuclides released into the atmosphere from Beloyarsk nuclear power plant with fast-neutron reactor for nine years in a row, from 2008 to 2016. It has been shown that the main radionuclides throw out into the atmosphere from Beloyarsk nuclear power plant are beta-active radionuclides. Based on data releases of the RPA "Typhoon", it has been conclude that radiation situation become worse insignificantly; beside on the new reactor BN-800 was put in operation in 2016. Using Spencer-Fano's equation, it was carried out the summary spectrum of emitted radionuclides. On example of Beloyarsk nuclear power plant, it was considered a question about ability of remote detection of raised radioactivity in the atmospheric radioactive plume. It has been shown that it possible to detect raised radioactivity in the emission plume from Beloyarsk nuclear power plant.

Modular nuclear reactor for a land-based power plant and method for the fabrication installation and operation thereof

International Nuclear Information System (INIS)

Craig, E. R.; Blumberg, B. Jr.

1985-01-01

A self-contained modular nuclear reactor which can be prefabricated at a factory location, nuclear-certified at the factory, transported to a field location for final assembly and connection to a large-scale electric-power generating facility. The modular reactor includes a prefabricated nuclear heat supply module and a plurality of shell segments which can be assembled about the heat supply module and which provide a form for the pouring and curing of a cementatious biological shield about the heat supply module. The modular reactor includes passive shutdown heat removal systems sufficient to render the reactor safe in an emergency. A large-scale power plant arrangement is disclosed which incorporates a plurality of the modular reactors

Reactor Coolant Pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Reactor coolant pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Expert system for maintenance management of a boiling water reactor power plant

International Nuclear Information System (INIS)

Hong Shen; Liou, L.W.; Levine, S.; Ray, A.; Detamore, M.

1992-01-01

An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, which must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components

Advanced CANDU reactors

International Nuclear Information System (INIS)

Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

1988-12-01

AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

International Nuclear Information System (INIS)

Pan Xiren; Zhang Chen.

1988-01-01

The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

Decision-making support system based on MAS for emergency response to nuclear accidents of marine pressurized-water reactor

International Nuclear Information System (INIS)

Chen Dengke; Zhang Dafa; Jiang Wei; Chen Yonghong

2007-01-01

Emergency decision-making to Marine Pressurized-Water Reactor (MPWR) was severely restricted by the complex environment. To enhance the emergency decision-making ability of MPWR, reducing the effect of emergencies, an emergency Decision-making Support System (DSS) which based on Multi-agent System (MAS) was presented. In the system, the HLA/RTI was used as the support environment, and the structure and the Control Agent (SCA), Analyse Agent (AA), Countermeasure Agent (CA), Evaluation Agent (EVA) and Environment Agent (ENA) were designed. The MAS were with the characteristics of autonomy, reactivity and initiative, which were fully used in the system to make effective decision for emergencies. (authors)

Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

Energy Technology Data Exchange (ETDEWEB)

Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

2017-11-01

The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

International Nuclear Information System (INIS)

Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto

2017-01-01

The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10"1"3 combinations and 10"1"1 great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)