WorldWideScience

Sample records for breeder reactor safety

  1. Optimisation of safety parameters in fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Optimisation of safety parameters is an important aspect to be considered in the design of nuclear power plant and also becomes extremely important activity to be followed up during the commissioning and operating phases of the plant taking into account the operational feed back and review of incidental situations and available diversity and reliability. Otherwise, the spurious/ superfluous trips on the reactor besides affecting the availability of the plant, initiate plant transients causing stress for the plant equipment resulting in reduction of plant life. This activity has a significant role to play in attaining the maximum availability of the plant, without compromising safety. The study and evolution of optimisation process in fast breeder test reactor (FBTR); at Kalpakkam has been an interesting and rewarding experience

  2. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  3. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  4. Embattled breeder reactor

    International Nuclear Information System (INIS)

    A commercial fuel-cloning machine, a nuclear breeder reactor, is yet to produce electricity in the United States. It is expensive in capital and fuel costs, its fuel that must be reprocessed can become a link to nuclear weapons manufacture, and its safety is no greater than conventional nuclear reactors. The breeder has had on-again/off-again administrative support from Washington. Opponents worry about escalating costs and failure to develop alternatives like solar energy. Proponents say fossil-fuel depletion will eventually force long-term renewable resources such as the breeder anyway. Some who share parts of both views oppose present policy regarding the Clinch River Breeder demonstration plant specifically. The correct choices on breeder concept development and commercialization will be known in 2050. 3 figures

  5. Operating safety experience of fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Operational safety criteria for nuclear reactors are very stringent and it is essential to incorporate adequate inherent and engineered safety features in the design to ensure safe operation of the reactor. Commissioning and operation of FBTR, being first of its kind in India based on nuclear chain reaction maintained by fast neutrons and use of high temperature liquid sodium as coolant, was a challenging task. Safe operation of the reactor for the past 17 years with good performance of sodium systems and the indigenous plutonium rich carbide fuel, touching a burn up level of 100 GWd/t has underlined the high level of design and operation competence achieved

  6. Breeder reactors

    International Nuclear Information System (INIS)

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components

  7. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    International Nuclear Information System (INIS)

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter

  8. Safety requirements expected to the prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of 'Monju' based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as '1F accident') occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up 'Advisory Committee on Monju Safety Requirements' consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to the prototype FBR 'Monju' considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee. (author)

  9. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  10. Compendium of computer codes for the safety analysis of fast breeder reactors

    International Nuclear Information System (INIS)

    The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available

  11. Prototype fast breeder reactor main and safety vessel surfaces in-service inspection mobile robot

    International Nuclear Information System (INIS)

    Periodic inspection of Prototype Fast Breeder Reactor (PFBR) main vessel and safety vessel is important to assess their structural integrity and to take remedial measures, if needed. PFBR is a pool type reactor and a safety vessel is provided in the design, which envelops the main reactor vessel. As the reactor inside is inaccessible, inspection can only be performed from outside the main vessel. Division of Remote Handling and Robotics, Bhabha Atomic Research Centre (BARC) in collaboration with Indira Gandhi Centre for Atomic Research (IGCAR) is working on the design of a proto-type mobile robot that would do the inspection of reactor vessel surfaces through the annular gap using friction grip when the reactor is in shutdown condition. This mobile robot will be inserted through the access holes at the top of the reactor vault leading to the annular space and moves around the vessel, carrying visual camera, lighting system and ultrasonic testing modules as accessories and also positions and orients them to do ISI of main and safety vessel surfaces. The details of the configuration of the VENTURE, method of achieving mobility around the vessel for coverage and adaptation to the variation in annular gap and other salient design features required to perform the ISI are briefed in this paper. (author)

  12. Fast breeder reactor

    International Nuclear Information System (INIS)

    The fluid-cooled fast breeder reactor described includes an outer cylindrical boundary wall, a plurality of canless fuel elements and breeder material elements received within the boundary wall and being in an array therein forming a fissionable fuel zone and a breeder material zone coaxially surrounding the fissionable fuel zone, a coolant supply system for applying fluid coolant at uniform pressure to the entire cross section within the cylindrical boundary wall, and flow guide devices extending substantially horizontally and disposed at different levels one above the other within the breeder material zone which coaxially surrounds the fissionable fuel zone, means for elastically securing the flow guide devices at alternate levels within the breeder material to the boundary wall, the flow guide devices at the levels intermediate the alternate levels being spaced by an annular gap from the boundary wall. 7 claims, 7 drawing figures

  13. Safety and core design of large liquid-metal cooled fast breeder reactors

    OpenAIRE

    Qvist, Staffan Alexander

    2013-01-01

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cyc...

  14. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  15. Development of standards and investigation of safety examination items for advancement of safety regulation of fast breeder reactor

    International Nuclear Information System (INIS)

    The purposes of this study are to prepare the fuel technical standard and the structure and materials standard of fast breeder reactors (FBRs), and to develop the requirements in a reactor establishment permission. The objects of this study are mainly the Monju high performance core and a demonstration FBR. In JFY 2012, the following results were obtained. As for the fuel technical standard, the fuel technical standard adapting the examination of integrity of the FBR fuels was prepared based on the information and data obtained in this study. As for the structure and material standard, the investigation of the revised parts of the standard was carried out. And as for the examination of the safety requirements, safety evaluation items for the future FBR plant and the fission products to be considered in a reactor establishment permission were investigated and examined. (author)

  16. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO2 process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  17. Fast breeder reactor research

    International Nuclear Information System (INIS)

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  18. Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor

    International Nuclear Information System (INIS)

    The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam-Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1-0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ∼1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using 'CRAFT' software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by β factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be ≤1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement

  19. The development of breeder reactors in the US

    International Nuclear Information System (INIS)

    This article discusses the early history of breeder development in the US, the early history of the fast reactor in the US, changes during the Carter administration, and the development of LMFBR technology. Topics considered include the intermediate-energy plutonium breeder, the molten plutonium breeder, the aqueous homogeneous reactor, the molten-salt reactor, the liquid metal-fueled reactor, electronuclear breeding, the Experimental Breeder Reactor-I, the Experimental Breeder Reactor-II, the Enrico Fermi Reactor, a programmatic change to ceramic fuel, the South East Fast Oxide Reactor, the sodium void coefficient, the 1000-MWe studies of 1964, the 1000-MWe studies of 1967-1969, the FARET design, the Fast Flux Test Facility, the Clinch River Breeder Reactor (CRBR), the gas-cooled fast breeder, the light-water breeder, materials for cladding and duct walls, and reactor safety. It is pointed out that the Congress opposes the construction of the CRBR, while the Reagan administration strongly supports it

  20. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  1. Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor

    International Nuclear Information System (INIS)

    The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2012, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination conducted for the reactor establishment permission, development of the analysis codes, such as core damage analysis code, were carried out following the planned schedule. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

  2. Safety-Evaluation Report related to the construction of the Clinch River Breeder Reactor Plant. Docket No. 50-537

    International Nuclear Information System (INIS)

    The Safety-Evaluation Report for the application by the United States Department of Energy, Tennessee Valley Authority, and the Project Management Corporation, as applicants and owners, for a license to construct the Clinch River Breeder Reactor Plant (docket No. 50-537) has been prepared by the Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission. The facility will be located on the Clinch River approximately 12 miles southwest of downtown Oak Ridge and 25 miles west of Knoxville, Tennessee. Subject to resolution of the items discussed in this report, the staff concludes that the construction permit requested by the applicants should be issued

  3. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  4. Significance of coast down time on safety and availability of a pool type fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Plant dynamics studies for quantifying the benefits of flow coast down time. • Establishment of minimum flow coast down time required for safety. • Assessment of influence of flow coast down on enhancing plant availability. • Synthesis of thermo mechanical benefits of flow coast down time on component design. - Abstract: Plant dynamic investigation towards establishing the influence of flow coast down time of primary and secondary sodium systems on safety and availability of plant has been carried out based on one dimensional analysis. From safety considerations, a minimum flow coast down time for primary sodium circuit is essential to be provided to limit the consequences of loss of flow event within allowable limits. Apart from safety benefits, large primary coast down time also improves plant availability by the elimination of reactor SCRAM during short term power failure events. Threshold values of SCRAM parameters also need optimization. By suitably selecting the threshold values for SCRAM parameters, significant reduction in the inertia of pumping systems can be derived to obtain desirable results on plant availability. With the optimization of threshold values and primary flow coast down behaviour equivalent to a halving time of 8 s, there is a possibility to eliminate reactor SCRAM during short term power failure events extending up to 0.75 s duration. Benefits of secondary flow halving on reducing transient thermal loading on components have also been investigated and mixed effects have been observed

  5. Enhanced passive safety features against ATWS of fast breeder reactors with capabilities of MA incineration

    Energy Technology Data Exchange (ETDEWEB)

    Ninokata, Hisashi; Sawada, Tetsuo; Sato, Manabu [Tokyo Institute of Technology (Japan)] [and others

    1997-12-01

    The paper gives an outline of the general and simple reactivity correlation method to identify the region of the major design parameters that assures power stabilization and passive shutdown of sodium-cooled large fast reactors under ATWS conditions. Based on the model developed, general design guidelines are shown that enhance passive capabilities being aimed at preventing sodium boiling and fuel failures in the events of ULOF and UTOP. Discussions extend to the influences of minor actinides loading in the core onto the passive safety features. 6 refs., 1 fig., 1 tab.

  6. Uncertainty evaluation of reliability of safety grade decay heat removal system of Indian prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Uncertainty analysis of failure frequency of SGDHRS of a medium sized fast reactor is studied. • Lognormal distribution of failure rate of components is taken with error factor of 3. • The error factor in the distribution of failure frequency in most cases is 3. • The relative importance of the safety components is brought out. - Abstract: Deterministic and probabilistic safety assessment of nuclear power reactor technology is very important in assuring that the design is robust and safety systems perform as per requirement. The parameters required as input data for such analysis have uncertainties associated with them. Their impact is to be assessed on the results obtained for such analyses and it affects the overall decision making process. Safety Grade Decay Heat Removal System (SGDHRS) is one of the safety systems in fast breeder reactors and itremoves decay heat after reactor shutdown. It is a critical safety system; hence failure frequency for SGDHR is targeted to be less than 1.0 × 10−7 per reactor year. By bringing diversity in some of the components of SGDHRS, such as sodium-to-sodium decay heat exchanger (DHX), sodium to air heat exchanger (AHX) and valves, one can achieve the targeted low failure frequency of SGDHRS. We perform uncertainty analysis of the reliability of such SGDHRS here. Uncertainty in failure rate (of components of SGDHRS) is assumed to follow the log-normal distribution with error factor of three. Monte Carlo method of sampling is used in MATLAB environment. Results are obtained in terms of mean, median and standard deviation values of failure frequency. Percentile and confidence interval analysis of mean values are also obtained. These provide 95 and 98 percentile and confidence interval values of 98%, 99% and 99.8%. It is found that error factor of failure frequency of SGDHRS is found to be less than 3 in all the cases except the one in which DHX, AHX and Valves are designed with diversity in design. It is to

  7. International cooperation on breeder reactors

    International Nuclear Information System (INIS)

    In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

  8. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    International Nuclear Information System (INIS)

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases

  9. Fast breeder reactor

    International Nuclear Information System (INIS)

    This paper outlined the present status of FBR development in six countries and reviewed Japanese activities on FBR development. Joyo experimental FBR has accumulated a lot of technical data including irradiation tests of advanced fuels and was now long shut down due to the partial obstruction of rotating plug movement. Monju prototype FBR reactor experienced a sodium leakage in its secondary heat transfer system during performance tests in December 1995 and had been shut down until May 2010. Feasibility study on commercialized FBR cycle system ended in March 2006 and proposed the concept of commercialized FBR cycle technologies. In order to plan a demonstration reactor, research and development of innovative technologies are conducted as the FaCT (Fast Reactor Cycle Technology Development) Project. In connection with the results of this research and development, a 5-party council of Japan was established to discuss processes of demonstration and commercialization of FBR cycle systems in Japan. Joint efforts were made for a demonstration reactor to be committed in 2015, in addition to start operation around 2025 aiming at the commercialization of FBR before 2050. (T. Tanaka)

  10. The Role of Energetic Mixed-Oxide-Fuel-Sodium Thermal Interactions in Liquid Metal Fast Breeder Reactor Safety

    International Nuclear Information System (INIS)

    Recent efforts dealing with the consequence assessment of low-probability core-disruptive accidents (CDAs) in liquid-metal fast breeder reactors (LMFBRs) suggest that unrealistic physical processes must be postulated in order to achieve energetic prompt burst conditions leading to a true hydrodynamic disassembly of the reactor core. Such calculations are, however, being used in the licensing process in order to provide an estimate of safety margins provided by a given design. Figure 1 illustrates calculations for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR), where the prompt critical excursion and associated ramp rates are induced by postulating various amounts and rates of collapsing fuel in a largely molten core (recriticality accident), and the mode of energy release considered is the expansion of fuel vapor resulting in sodium-slug impact on the reactor vessel head. The VENUS-II code is used to calculate the disassembly motion and power histories during disassembly Elementary thermodynamic calculations provide the source term based upon expansion of the fuel from an initial temperature distribution specified by VENUS calculations, and the REXCO series of codes provide a hydrodynamic calculation of the pressure propagation coupled with an analysis of the structural response of the important system components. The work potential resulting from fuel collapse and hydrodynamic disassembly is very sensitive to small variations in the ramp rate. Since material motions associated with postulated conditions leading to energetic prompt critical excursions cannot be described with sufficient accuracy to provide reasonable bounds on ramp rates, an adequate margin of safety with current design is difficult to claim if these conditions cannot be ruled out. This implies that in addition to coherent gravity collapse, the possibility of pressure-driven (fuel-coolant interaction) collapse must be considered. Furthermore, the work potential

  11. 04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

    International Nuclear Information System (INIS)

    The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

  12. Liquid-metal fast-breeder reactors: Preliminary safety and environmental information document. Volume VI

    International Nuclear Information System (INIS)

    Information is presented concerning LMFBR design characteristics; uranium-plutonium/uranium recycle homogeneous core; uranium-plutonium/uranium spiked recycle heterogeneous core; uranium-plutonium/uranium spiked recycle homogeneous core; uranium-plutonium/thorium spiked recycle heterogeneous core; uranium-plutonium/thorium spiked recycle homogeneous core; thorium-plutonium/thorium spiked recycle homogeneous core; denatured uranium-233/thorium cycle homogeneous core; safety consideration for the LMFBR; and environmental considerations

  13. The design of the Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India has a moderate uranium reserve and a large thorium reserve. The primary energy resource for electricity generation in the country is coal. The potential of other resources like gas, oil, wind, solar and biomass is very limited. The only viable and sustainable resource is the nuclear energy. Presently, Pressurised Heavy Water Reactors utilizing natural uranium are in operation/under construction and the plutonium generated from these reactors will be multiplied through breeding in fast breeder reactors. The successful construction, commissioning and operation of Fast Breeder Test Reactor at Kalpakkam has given confidence to embark on the construction of the Prototype Fast Breeder Reactor (PFBR). This paper describes the salient design features of PFBR including the design of the reactor core, reactor assembly, main heat transport systems, component handling, steam water system, electrical power systems, instrumentation and control, plant layout, safety and research and development

  14. Analysis of unprotected transients with control and safety rod drive mechanism expansion feedback in a medium sized oxide fuelled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sathiyasheela, T., E-mail: sheela@igcar.gov.in; Natesan, K.; Srinivasan, G.S.; Devan, K.; Puthiyavinayagam, P.

    2015-09-15

    Highlights: • Possibilities of enhancing safety under ULOF and UTOP accidents. • CSRDM expansion feedbacks under unprotected transients. • CSRDM expansion feedback enhances the safety of fast reactors. • CSRDM expansion feedbacks ensuring enough time for initiating safety actions. - Abstract: Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000 s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000 s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis.

  15. A report on (interim) evaluation of research and development subjects in fiscal year 2000. Evaluation subject on the 'Safety research in fast breeder reactor'

    International Nuclear Information System (INIS)

    Safety research as a basis R and D supporting development of the fast breeder reactor (FBR) has been practiced at aims of development, admittance and operation/maintenance of a fast experimental reactor, 'Joyo' and a fast breeder prototype reactor, 'Monju' and of reflection to a proof reactor plan promoted by the electric utility. However, at present, in order to reflect FBR cycle actual use strategy survey research, decision of importance in research is promoted to effectively reflect their research results to judgment and investigation on consistency of various candidate concepts. Here was carried out on some evaluations on research program and practicing method of coming five years on conventional research results, reflection to the second period of the actual use strategy survey research, and practice of national safety research yearly plan at a center of past five years on contribution to FBR development and safety regulation in Japan. Here were described on aim and meaning of the R and D, establishment of target, planning, practicing system, and results. (G.K.)

  16. Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor (Annual safety research report, JFY 2010)

    International Nuclear Information System (INIS)

    The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2010, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination achieved for the reactor establishment permission, development of the analysis codes such as core seismic analysis code, core safety analysis code and core damage analysis code were carried out according to the plan. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied, and the seismic PSA to evaluate residual risk was studied. (author)

  17. Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor (Annual safety research report, JFY 2011)

    International Nuclear Information System (INIS)

    The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2011, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination achieved for the reactor establishment permission, development of the analysis codes such as core seismic analysis code, core safety analysis code and core damage analysis code were earned out according to the plan. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied, and the seismic PSA to evaluate residual risk was studied. (author)

  18. The nuclear question at the start of the '80s: the breeder reactor

    International Nuclear Information System (INIS)

    The four newspaper articles and the letter cover the following matters: general introduction about breeder reactors and the situation in Swedish politics; visit to Dounreay to discuss breeder reactors (breeding, safety, plutonium production, radiation protection); PuO2-UO2 mixed fuel; description of breeder reactors; efficiency in use of U-235; DFR and PFR; breeder reactors in Swedish politics (arguments for and against nuclear power in general, breeder reactors in particular); discussion of the future of nuclear power in Sweden. (U.K.)

  19. The United States of America fast breeder reactor program

    International Nuclear Information System (INIS)

    The reasons for the development of the fast breeder reactor in the United States are outlined, and the LMFBR program is discussed in detail, under the following headings: program objectives, reactor physics, fuel and materials development, fuel recycle, safety, components, plant experience program (Near Commercial Breeder Reactor). The special facilities to be used at each stage of the program are described. It is planned that the Near Commercial Breeder Reactor will be complete in 1986, and commercial plants should follow in rapid succession. An alternate fast reactor concept (Gas Cooled Fast Reactor) is outlined. The Environmental Impact Statement for the proposed program is summarized, and the cost benefit analysis supplied as part of the Environment Statement is also summarized. (U.K.)

  20. Exploring new coolants for nuclear breeder reactors

    International Nuclear Information System (INIS)

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F2Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F2Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F2Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void reactivity coefficients

  1. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  2. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  3. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

    Results of a conceptual design study of a 233U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed

  4. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  5. Future designs of breeder reactors (Europe, USA)

    International Nuclear Information System (INIS)

    Sodium-cooled reactors with a fast neutron core today are the only fission reactors that offer the reactor physics required for the breeding process and the complete conversion of U-238 or Th-232 into fissile fuel. There are currently five prototype breeder reactors in operation in England, France, and the USSR. The trends observable in development work aim at reducing capital cost, enhancing and improving passive shutdown performance, and simplifying the fuel cycle. (orig.)

  6. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  7. Status of fast breeder reactor development in the United States

    International Nuclear Information System (INIS)

    support on a schedule consistent with an internationally developed cooperative program; fuel cycle - develop a reprocessing technology base emphasizing fast breeder systems but generic to all reactor types. In addition, the US Nuclear Regulatory Commission's Advanced Reactor Safety Research Program provides research to support licensing evaluations for demonstration and prototype plants. The United States Government is also actively involved in cooperative breeder research and development activities with other nations and international organizations

  8. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  9. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors)

  10. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  11. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  12. A fast breeder reactor development scheme for Brazil

    International Nuclear Information System (INIS)

    Fast breeder reactors will be necessary in the next century in order to meet increasing demands for electricity resulting from industrialization and general improvement of standards of living. A scheme for the development of liquid metal fast breeder reactors in Brazil is proposed. Emphasis are placed on reactor safety in order to promote public acceptance, on utilization of thorium that is abundant in the country, and on consistency and smoothness of the development. The initial step is the construction and operation of a 5 MW experimental fast reactor in order to acquire basic experiences and technologies. The second step is the construction of a series of small power plants which should assure a ssound technological development. The reactor is designed with particular emphasis on safety and ease of operation. Demonstration of safety and reliability with small units would enhance public acceptance. In the final phase, when fast breeder reactors are to play a central role in electricity generation, large power plants that utilize both uranium and thorium fuel cycles will be built to establish a practically permanent power system. (Author)

  13. Status of the DEBENE fast breeder reactor development, March 1979

    International Nuclear Information System (INIS)

    Status report of the Fast-breeder reactor development in Germany covers the following: description of the political situation in Federal republic of germany during 1978; international cooperation in the field of fast reactor technology development; operation description of the KNK-II fast core experimental power plant; status of construction of the SNR-300; results of the research and development programs concerned with fuel element, cladding, absorber rods and core structural materials development; sodium effects; neutron irradiation effects on SS properties; reactor physics related to experiments in fast critical assemblies; fast reactor safety issues; core disruption accidents; sodium boiling experiments, measuring methods developed; component tests

  14. History and evolution of the breeder reactor

    International Nuclear Information System (INIS)

    The concept of the breeder reactor is almost as old as the idea of the nuclear reactor itself. From the very first years following the discovery of nuclear fission, scientists and technicians tried to turn mankind's eternal dream into reality; that is, enjoy an abundant source of energy without using up our raw material reserves. Nuclear energy offered several solutions to realize this dream. One of them, fusion, seemed out of our grasp in the near future. But fission of 235U was possible, and the Manhattan Project soon furnished ample proof of this theory. However, everyone working in this field was conscious of the fact that thermal neutron reactors make very inefficient use of the energy potential contained in natural uranium. The solution was to use in a core sufficiently rich in fissile matter, the excess neutrons to convert the 238U, so poorly used by other types of reactors, into fissile 239Pu. Regeneration, or 'breeding' of fuel, can multiply the energy drawn from a ton of uranium by a factor of 50 to 100. This would enable us to ward off the specter of an energy shortage and the rapid depletion of uranium mines. As early as 1945 in Los Alamos, Enrico Fermi stated: 'The country which first develops a breeder reactor will have a great competitive edge in atomic energy.' The development of the breeder reactor in the USA and around the world is discussed

  15. Improved structural materials for fast breeder reactors

    International Nuclear Information System (INIS)

    Electricity plays a crucial role in the economic development of our country. Coal is the primary fuel for generation of electricity in India as in many other countries. In India, generation of power by nuclear reactors is very important because of (i) availability of large thorium resource, (ii) constraints on setting up of fossil fuel based power plants and (iii) the negligibly small green house gas emissions by nuclear energy. The nuclear programme of the country is being implemented in three stages: (i) pressurized heavy water reactors of the CANDU type, (ii) sodium-cooled fast reactors and (iii) thorium-based reactors. Sodium-cooled fast reactor (SFR) technology is envisioned to make use of the large thorium reserves available. India has undertaken and made rapid strides in developing SFR technology and building of fast reactors for energy generation. A Fast Breeder Test Reactor (FBTR) of 40 MWt is operating successfully for over 25 years at Indira Gandhi Centre for Atomic Research. Based on the design, construction and operational experience, a 500 MWe Prototype Fast Breeder Reactor (PFBR) has been designed indigenously and is in an advanced stage of construction. Its design is being further optimised for enhanced economy with respect to cost of electricity production, for use in commercial reactors. Currently, several R and D programmes are under implementation for the development of new materials required for improved economy of commercial fast reactors

  16. Binary breeder reactor: an option for Brazilian energy future

    International Nuclear Information System (INIS)

    To assure a continued supply of electric energy beyond a few decades from now, developmemnt of fast breeder reactors is a necessity. Binary fueled LMFBRs combine an improvement in the inherent safety of fast reactors and an efficient use of the abundant thorium. A nuclear system that starts with PWRs and gradually shifts to a FBR system or to a FBR-PWR symbiotic system appears to be the most reasonable one for Brazil. Natural uranium requirements and possible sequences of reactor introductions are discussed for some postulated Brazilian situations. A permanent system of approx. 100 GWe capacity can be established based on the estimated resource of natural uranium. (Author)

  17. Binary breeder reactor an option for Brazilian energy future

    International Nuclear Information System (INIS)

    To assure a continued supply of electric energy beyond a few decades from now, development of fast breeder reactors is a necessity. Binary fueled LMFBRs combine an improvement in the inherent safety of fast reactors and an efficient use of the abundant thorium. A nuclear system that starts with PWRs and gradually shifts to a FBR system or to a FBR-PWR symbiotic system appears to be the most resonable one for Brazil. Natural uranium requirements and possible sequences of reactor introductions are discussed for some postulated Brazilian situations. A permanent system of approximatelly 100 GWe capacity can be established based on the estimated resource of natural uranium. (Author)

  18. Status of fast breeder reactor development in the United States

    International Nuclear Information System (INIS)

    The energy policy of the United States is aimed at shifting as rapidly as practicable from an oil dependent economy to one that relies heavily on other fuels and energy sources. Nuclear power Is now and is expected to continue to be an important factor in achieving this goal. If nuclear power is to contribute to a solution of future energy needs, demonstration of the breeder reactor as a viable source of essentially inexhaustible energy supply is essential. The US DOE program for development of the fast breeder reactor has witnessed some notable events in the past year. Foremost among these Is the successful operational testing of the Fast Flux Test Facility (FFTF), located at.the Hanford Engineering Development Laboratory. The reactor reached full design power of 400 MW(t) on December 21, 1980, and has performed remarkably close to design specifications. Design of the Clinch River Breeder Reactor Plant (CRBRP), a 375 MW(e) LMFBR, is now over 80 percent complete. About $530 million in components have been ordered; component deliveries total approximately $124 million; work-in-process totals another $204 million. Construction of the plant, however, has been suspended since 1977. With the concurrence of the U.S. Congress and approvals from the appropriate authorities work on the safety review and site clearing for construction can resume. The Conceptual Design Study for a large, 1000 MW(e) LMFBR Large Developmental Plant was recently completed on a schedule commensurate with submission of a full report to the Congress at the end of March, 1981. This report is the culmination of a study which began in October, 1978 and involved contributions from U.S. reactor manufacturers and US DOE laboratories. The US DOE is carrying forward a comprehensive technology development program. This effort provides direct support to the FFTF and CRBRP projects and to the LDP. It also supports technology development which is generic to the overall LMFBR program. Funding for breeder

  19. An Evaluation of liquid metal leak detection methods for the Clinch River Breeder Reactor Plant

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C.J.; Doctor, S.R.

    1977-12-01

    This report documents an independent review and evaluation of sodium leak detection methods described in the Clinch River Breeder Reactor Preliminary Safety Analysis Report. Only information in publicly available documents was used in making the assessments.

  20. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  1. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  2. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  3. Alternate fuel cycles for fast breeder reactors

    International Nuclear Information System (INIS)

    In this contribution to the syllabus for Subgroup 5D, a full range of alternate breeder fuel cycle options is developed and explored as to energy supply capability, resource utilizations, performance characteristics and technical features that pertain to proliferation resistance. Breeding performance information is presented for designs based on Pu/U, Pu/Th, 233 U/U, etc. with oxide, carbide or metal fuel; with lesser emphasis, heterogeneous and homogeneous concepts are presented. A potential proliferation resistance advantage of a symbiotic system of a Pu/U core, Th blanket breeder producing 233 U for utilization in dispersed LWR's is identified. LWR support ratios for various reactor and fuel types and the increase in uranium consumption with higher support ratios are identified

  4. The fast breeder reactor. v. 1

    International Nuclear Information System (INIS)

    The Energy Committee's report was prepared after hearing evidence (the minutes of which are published in Volume II) from the Central Electricity Generating Board, the United Kingdom Atomic Energy Authority and the Department of Energy. Memoranda received from other interested bodies or individuals were also considered and members of the Committee visited fast breeder projects in France, West Germany and Japan. As well as the development of the fast reactors, the economics and timescale were reviewed. The particular case of the fast breeder reactor and proposed fuel reprocessing plant at Dounreay was considered. The main conclusion is that major expenditure on fast reactor programmes can only be justified if there is a potential economic case, i.e. if the fuel cycle costs are lower than for PWRs. This would only be the case if uranium costs increased greatly. It is not considered worthwhile to participate in the European Fast Reactor although this should be reviewed in 1993 and 1997. The Committee agree with the Government's decision to cease funding the PFR in 1994 and endorses the need to regenerate the local economy which will be affected by this decision. (UK)

  5. Large scale breeder reactor pump dynamic analyses

    International Nuclear Information System (INIS)

    The lateral natural frequency and vibration response analyses of the Large Scale Breeder Reactor (LSBR) primary pump were performed as part of the total dynamic analysis effort to obtain the fabrication release. The special features of pump modeling are outlined in this paper. The analysis clearly demonstrates the method of increasing the system natural frequency by reducing the generalized mass without significantly changing the generalized stiffness of the structure. Also, a method of computing the maximum relative and absolute steady state responses and associated phase angles at given locations is provided. This type of information is very helpful in generating response versus frequency and phase angle versus frequency plots

  6. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Liquid Metal Cooled Fast Breeder Reactors (LMFBR) require mechanisms operating in various sodium liquid and sodium vapor environments for extended periods of time up to temperatures of 900 K under different chemical properties of the fluid. The design of tribological systems in those reactors cannot be based on data and past experience of so-called conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, operating conditions specific to nuclear reactors and prevailing especially in the nuclear part of such facilities pose special problems. Therefore, in the framework of the R and D-program accompanying the construction phase of SNR 300 experiments were carried out to provide data and knowledge necessary for the lay-out of friction systems between mating surfaces of contacting components. Initially, screening tests isolated material pairs with good slipping properties and maximum wear resistance. Those materials were subjected to comprehensive parameter investigations. A multitude of laboratory scale tests have been performed under largely reactor specific conditions. Unusual superimpositions of parameters were analyzed and separated to find their individual influence on the friction process. The results of these experiments were made available to the reactor industry as well as to factories producing special tribo-materials. (orig.)

  7. Safety-Evaluation Report related to the construction of the Clinch River Breeder Reactor Plant. Docket No. 50-537

    International Nuclear Information System (INIS)

    The purpose of this appendix is to describe the staff's evaluation of hypothetical core disruptive accidents which, for analytical purposes, have been postulated to occur in the CRBR. This introduction is divided into three major parts. The first background information. The second provides an overview of potential CDA initiating events and consequences considered for the CRBR. The third describes the guidelines used in evaluating CDAs for the CRBR. A schematic view of major components of the reactor systems is provided. The staff's evaluation of the major areas associated with the assessment of CDAs is presented

  8. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232Th and 233U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239Pu and 233U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235U and nuclear waste management issues. A mixture of thorium and uranium (232Th + 233U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239Pu. Initial feed of 233U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233U a theoretical evaluation has been used to derive the data for the source of 233U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233U as an alternate to 235U and an alternate use for light water reactor used fuel as a blanket for

  9. Reactor safety engineering

    International Nuclear Information System (INIS)

    The concept of the work is such that the basic safety philosophy for nuclear power plants as well as the safety features of both types of light water reactors, pressurized and boiling water reactors, and of the fast breeder reactor are dealt with. With the pressurized and boiling water reactors also variations, due to different supplies are mentioned. The state of development considered is characterized by the results of the American reactor safety study having very much influenced the way of presentation and the validity of the information contained. In the introduction the attentive reader is made familiar with the basic traits of safety engineering, the traditional deterministic way of proceeding being supplemented by a detailed illustration of probabilistic means used in the safety analysis. Added to this are comparative descriptions of the individual safety features, their design and mode of operation. There are, e.g., detailed discussion of the emergency core cooling systems, the power supply systems, the reactor protection system, and the containment. Special chapters are attributed to transients with and without the fast shutdown system working and to loss of coolant. The so-called external events are treated somewhat shortly whereas much space is given to core melting problems. The treatment of important events from the safety point of view, including the section on Harrisburg added for reasons of immediate interest, is limited to phenomenological description. (orig.)

  10. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  11. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed

  12. Safeguards in Prototype Fast Breeder Reactor Monju

    International Nuclear Information System (INIS)

    The assemblies loaded in the core and stored in the ex-vessel storage tank (EVST) are in liquid sodium in the Japanese prototype fast breeder reactor (FBR) Monju. Since it is difficult to apply a direct verification procedure for the fuel assemblies in these areas, a dual containment and surveillance system consisting of two monitoring devices such as surveillance camera and radiation monitor that are functionally independent has been applied. In addition, the Monju Remote Monitoring System was developed to strengthen the continuous surveillance and to reduce the load of the inspection activities. Furthermore, the ex-vessel transfer machine radiation monitor (EVRM) and the exit gate monitor (EXGM) were upgraded to strengthen the monitoring of spent blanket fuel assemblies and to improve the reliability of distinguishing between fuel assemblies and non-fuel items. As the result, the integrated safeguards was introduced in November 2009, and the effective safeguards activities have been implemented in Monju. (author)

  13. Calculations of two-phase flows in the liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Mathematical models used for the safety analysis of liquid metal cooled fast breeder reactors are considered. Models, taking into account sodium boiling in reactor channels (one-dimensional and many-dimensional approaches), fuel cladding melting, and movement of molten materials during loss of coolant, accidents are described

  14. On the development of fast breeder reactors and the use of thorium in Brazil

    International Nuclear Information System (INIS)

    This work presents a discussion on the possibility of construction of fast breeder reactors in Brazil. It is specially concerned with the use of thorium which is abundant in our country. The main advantages of this projects are: develop fuel and reactor technology in Brazil, increase thorium research, demonstrate the safety of LMFBR and promote its public acceptance. (A.C.A.S.)

  15. Reprocessing of fast breeder reactor fuels in France

    International Nuclear Information System (INIS)

    The reprocessing of breeder reactor fuels is a direct technical descendant of the reprocessing of thermal reactor fuels which was developped first. The process used is in both cases the PUREX process, which consists in dissolution by nitric acid followed by selective extraction using TBP. In France, the application of this technique to breeder reactor fuels greatly benefited from the scientific and industrial experience initially acquired with metallic fuels of the MAGNOX type and then with oxide fuels of the LWR type

  16. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    International Nuclear Information System (INIS)

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and 233U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles

  17. Innovations in Equipment Erection of Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is sodium cooled, pool type reactor with generating capacity of 1250 MWt/500 MWe. Reactor assembly consists of large dimensional vessels like Safety vessel (13.54 m diameter, 12.8 m height and weight approximately 155 MT) and Main vessel (12.9 m diameter, 12.94 m height and weight approximately 202 MT including core catcher, core support structure and cooling pipes) and Steam generator (26 m length, 1.5 m diameter, and weight approximately 35 MT). PFBR reactor equipment erection was a challenging task where thin walled vessels had transported and handled with utmost precaution to avoid radial forces on the vessels which could buckle the vessels. There was a real challenge in lifting the vessels without swing, placement of large size and heavy vessel at a distance of 57 m where the crane operator had no line of site to the equipment being erected. To handle such over dimensional reactor components many mock-up tests had been carried out before erection and gained lot of confidence. Lot of care had been taken during lifting, handling and erection of thin walled over dimensional reactor components with innovative methods used for lifting fixtures, guiding arrangements, alignment fixtures and achieved the stringent erection tolerances. This paper discusses the first ever experiences gained during the handling and erection of such thin walled, over dimensional reactor components at PFBR site. (author)

  18. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    The reactor concept analyzed is a 233UF6 core surrounded by a molten salt (Li7F, BeF2, ThF4) 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

  19. UF6 breeder reactor power plants for electric power generation

    Science.gov (United States)

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

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (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.

  20. Operating experience of fast breeder reactors in the USSR

    International Nuclear Information System (INIS)

    The operating experience results of BN-600, BN-350, BOR-60 and BR-10 fast breeder reactors are presented. The fast reactors design and operation experience in the USSR has demonstrated their high operational qualities, safety, reserves of improvement. After 11 years' operation the BN-600 and 18 years' operation the BN-350 these two nuclear plants present a very satisfactory global loading rate of above 65%. The operation flexibility of the nuclear power plants and, in particular, the possibility of operation at 2/3 nominal power (BN-600) and at 4/5 and/or 3/5 nominal power (BN-350) have allowed for these loading rates to be reached in spite of numerous steam generators and pumps replacement. (J.P.N.)

  1. Fast-breeder-power reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology. The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc. (author)

  2. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

  3. The breeder reactor in electricity supply

    International Nuclear Information System (INIS)

    Forecasts are made of Britain's energy prospects in the year 2000. It is concluded that fossil fuels and renewable energy sources will leave an energy gap and that dependence on nuclear power will be substantial. There will, however have been a rapid depletion of readily available uranium ore reserves and a growing availability of plutonium from thermal reactors. Britain's resources of plutonium and depleted uranium which the fast breeder reactor can use - will equal many thousand million tonnes of coal. Our nuclear programme should therefore include one or two FBRs. Resources should be pooled internationally and plants built to prove alternative options and consolidate an already highly developed technology. In Britain our earliest nuclear (Magnox) stations have served as well. In Scotland, where next year an estimated 30% of electricity output will be nuclear, Hunterston 'B' AGR has had a splendid first year of operation, and pumped storage capacity in Scotland has extended the benefits of low-cost generation. The FBR has many very satisfactory engineering features and is eminently controllable and well behaved. It is compact, with relatively low cooling-water requirements and it could be built, one hopes, close to our load centres. There can be confidence that it will be proved safe. An order for an FBR, on the earliest timescale that fits in with evidence of successful operation of the Dounreay PFR and an agreed international programme, would serve the national interest and ensure the survival of our plant manufacturers, so badly hit by the effects of stagnation of the U.K. economy. (author)

  4. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Aiming to assure the continued utilization of fission energy, the development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed, as the best type for the Brazilian nuclear system in the future. The inherent safety characteristics are superior to current fast breeder reactors and an efficient utilization of thorium can be realized. The construction and operation of an experimental fast reactor is the first step and a basic tool for the development of FBRs technologies. A serie of core design for an 90 MW FBR is studied and the possible options and sizes of the main parameters are identified. (E.G.)

  5. Operation and maintenance experience with control rod and their drive mechanisms of fast breeder test reactor

    International Nuclear Information System (INIS)

    This paper explains the functional and construction features of Control Rod Drive Mechanism (CRDM) and control rod used in Fast Breeder Test Reactor (FBTR) which is a 40 MWt loop type sodium cooled fast reactor. It discusses all safety related incidents and failures encountered during its service in reactor, the solutions evolved and modifications carried out to prevent recurrence. It also details the maintenance activities and periodical surveillance carried out. The results of a reliability analysis done are also discussed. (author)

  6. Italian position paper on the safety analysis of liquid metal fast breeder reactors as related to sodium fires. The PEC reactor

    International Nuclear Information System (INIS)

    To obtain a deep understanding of physical phenomena and engineering problems connected to sodium fires, and to optimize the utilization of human and financial resources available, CNEN (now ENEA) has decided to join the French Commissariat a l'Energie Atomique (CEA) in the realization of a Franco-Italian experimental programme on sodium fires, named ESMERALDA. As for design preventions for PEC reactor (a fast flux, liquid metal cooled, fuel element testing reactor) fundamental choices were made taking into account all available knowledge, but with particular reference to the results of CEA's previous experiments on sodium fires. More detailed design analysis will be possible in the future, based on experimental results coming from the ESMERALDA programme

  7. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

  8. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  9. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    International Nuclear Information System (INIS)

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option

  10. Design evaluation system for class 1 component of fast breeder reactor plants

    International Nuclear Information System (INIS)

    The development of a new type of nuclear power plant called Fast Breeder Reactor has been greatly promoted of late by the Power Reactor and Nuclear Fuel Development Corporation (PNC) and others in hopes of replacing Light Water Reactors so far prevailing in Japan. Fast Breeder Reactor, unlike Light Water Reactor, is subjected to elevated temperature within the creep temperature range for long duration, thus requiring higher structural standards for reliability as well as for safety. In this connection, PNC has been conducting many years' research and development to establish reliable design methods based on an advanced analysis taking into consideration elevated temperature properties of materials, and finally worked out Structural Design Guide for Class 1 Components of the prototype of Fast Breeder Reactor in elevated temperature service. The POST-DS system in this paper has been developed as an design evaluation system based on the above design guide, by Mitsui Engineering and Shipbuilding Co., Ltd. since 1979 in accordance with a commission given by PNC. Using the results of the heat transfer analysis and stress analysis for Class 1 Components of Fast Breeder Reactor, this system can evaluate the following factors. 1) Primary stress limit, 2) Strain limit, 3) Creep Fatigue damage. (author)

  11. Status of fast breeder reactor development in Germany

    International Nuclear Information System (INIS)

    The KNK, the sodium cooled compact reactor is an experimental nuclear power plant of 20 MW electric power. Since 1977, it has been operated with fast reactor cores as KNK II. The KNK II/3 core was designed. The core fabrication has been largely completed. In 1990, the KNK II plant achieved a time availability of 56%. On January 8, 1991 KNK II was shut down for inspection. Since pre-nuclear commissioning was completed the Kalkar Nuclear Power Station SNR 300 has been operated in a mode similar to that of a power station. In March 1991 the financing partners decided not to prolong the standby phase because they do not think that the last construction permit and the operation permit will be issued within a definite period of time. The partners were convinced that the lack of progress in the licensing procedure was not caused by basic safety deficiencies of the project but by the way the licensing procedure was executed. The German fast breeder programme is now concentrated on contributions to the European Fast Reactor. (author)

  12. Super-Phenix: first step in the commercial development of breeder reactors

    International Nuclear Information System (INIS)

    Super-Phenix is an indispensable step in a program for the development of the breeder reactors. Its power and technical options are recalled. The administrative directives -particularly concerning safety- are in agreement with the regulation enforced in France relative to nuclear power stations and work has begun on the Creys-Malville site within the international framework where the development of breeders in Europe now finds its place. The saving in uranium due to breeders is certain, but their cost prices remain to be proved. The construction of Super-Phenix is the proof that it is possible to rely on the breeders in the final energetic results for the future, but it is of a type which can still be modified to a considerable extent, especially in order to reduce investments

  13. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    In the power increase performance test of the experimental fast reactor ''Joyo'', which was in progress since April, the first stage of the rated thermal output of 50 MW has been accomplished on July 5. Thereafter, the continuous opeation test at 50 MW for 100 hours was performed for the verification of its overall operational performance from August 13 to 16. The safety evaluation for power increase up to 75 MW and 100 MW, which was under way since September, last year, was completed, and the power increase was licensed on September 20. Concerning the design of the prototype fast breeder reactor ''Monju'', the studies on the specifications of the Construction Preliminary Design (2) have been finished. In respect of the analysis and preparation of materials for the Safety Licensing by the Committee, the developments of the analytical codes for rupture propagation in the heat transfer tubes of steam generators and for decay heat have been conducted. In the construction site surveys, the third geological structure survey and beach deformation survey have all ended, while the meteorological and seismic observations, the prediction of the diffusion of drained warm water, the survey of river flow, etc. are now under way. A report on the survey conducted on the construction site in Shiraki was received by the Fukui prefectural government in July, and the copies of a report on the assessment of environmental effect were submitted in August to both the national government and the Fukui prefectural government. The situations of progress of the research and development works on reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety and steam generators are reported. (Nakai, Y.)

  14. Safety evaluation report related to the construction of the Clinch River Breeder Reactor Plant. Docket No. 50-537. Suppl. 1

    International Nuclear Information System (INIS)

    Since the preparation of the Safety Evaluation Report the Advisory Committee on Reactor Safeguards considered the Clinch River construction permit license application at its 276th meeting and subsequently issued a favorable report, dated April 19, 1983 to the Commission (See Appendix I of this report). Additional documents associated with the application have been reviewed and a number of meetings have been held with the applicants. These events and documents are identified in Appendix E to this supplement. This supplement, SSER-1, to the Safety Evaluation Report, provides an evaluation of additional information received from the applicants since preparation of the SER regarding previously identified outstanding review items, and our response to the comments made by the Advisory Committee on Reactor Safeguards in its report

  15. Status of fast breeder reactor development in the United States of America - April 1984

    International Nuclear Information System (INIS)

    The Breeder Technology program continues to produce viable information on fuel performance, nuclear systems technology, and power conversion technology. The unique testing capabilities design into the FFTF have resulted in well-validated materials and fuels irradiation information that has confirmed and extended previous data bases. Current directions for the research and development program are to improve the technology for power conversion systems, components, instrumentation, and materials technology to the point where cost reduction and reliability potentials are realized. Operation of the breeder test facility complex at the Hanford Engineering Development Laboratory (HEDL), the Energy Technology Engineering Center (ETEC), and the Argonne National Laboratory (ANL) continues to provide the experience base and test capability for the breeder R and D effort. International cooperation will be even more important in the future than in the past for several reasons. Significant new investments still have to be made in breeder R and D to improve designs, achieve economic competitiveness and to develop practical breeder fuel cycle capabilities. Progress can be accelerated, redundancies avoided, and economics achieved if nations coordinate their programs, and where possible, divide up the work. In addition, there is clear mutual benefit in encouraging the countries involved in breeder development to harmonize standards and regulations related to safety. It is also important that the advanced nations work together closely in assuring that adequate international safeguards, export controls, and national physical security measures keep pace with breeder reactor and fuel cycle developments

  16. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  17. Clinch River Breeder Reactor Plant Project: construction schedule

    International Nuclear Information System (INIS)

    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

  18. Symposium on key questions about the fast breeder reactor

    International Nuclear Information System (INIS)

    Except for several introductions on various aspects of the fast breeder reactor development this paper contains the full texts of the discussions held in the sub-groups panels on resp. technical matters, environment and health, society, politics and economics. The main issues of each discussion are summarized

  19. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  20. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  1. The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    This report discusses the Shippingport Atomic Power Station, located in Shippingport, Pennsylvania, which was the first large-scale nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. A program was started in 1953 at the Bettis Laboratory to confirm the practical application of nuclear power for large-scale electric power generation. It led to the development of zirconium alloy (Zircaloy) clad fuel element containing bulk actinide oxide ceramics (UO2, ThO2, ThO2 -- UO2, ZrO2 -- UO2) as nuclear reactor fuels. The program provided much of the technology being used for design and operation of the commercial, central-station nuclear power plants now in use. The Shippingport Pressurized Water Reactor (PWR) began initial power operation on December 18, 1957, and was a reliable electric power producer until February 1974. In 1965, subsequent to the successful operation of the Shippingport PWR (UO2, ZrO2 -- UO2 fuels), the Bettis Laboratory undertook a research and development program to design and build a Light Water Breeder Reactor (LWBR) core for operation in the Shippingport Station. Thorium was the fertile fuel in the LWBR core and was the base oxide for ThO2 and ThO2 -- UO2 fuel pellets. The LWBR core was installed in the pressure vessel of the original Shippingport PWR as its last core before decommissioning. The LWBR core started operation in the Shippingport Station in the autumn of 1977 and finished routine power operation on October 1, 1982. Successful LWBR power operation to over 160% of design lifetime demonstrated the performance capability of the core for both base-load and swing-load operation. Postirradiation examinations confirmed breeding and successful performance of the fuel system

  2. Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

    International Nuclear Information System (INIS)

    Research highlights: → ULOF analysis of metal (U-Pu-6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR. → Uncertainties (typically 20%) on the sensitive feedback parameters. → Sensitive parameters - core radial feedback and sodium void reactivity effect. → Transient behavior of both 500 MWe and 1000 MWe core are benign under ULOFA. → For 1000 MWe inherent safety is assured with limited sodium void reactivity. - Abstract: Unprotected loss of flow (ULOF) analysis of metal (U-Pu-6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.

  3. Tritium dynamics in fusion reactor solid breeder

    International Nuclear Information System (INIS)

    In the field of the NET research progrm, the chemical and diffusive processes involved in solid ceramic breeder materials have been analysed. A mathematical model describing the phenomena has been developed to obtain a quantitative evaluation for a first design approach. The data obtained by means of the above mentioned model are in good agreement with the data obtained by other research groups working in Europe and in United States. The computer codes BLANKET2, MC2, FWBC, have been developed to simulate the phenomena

  4. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The twenty-second Annual Meeting of the International Working Group on Fast Reactors took place in Vienna, 18-21 April 1989. Nineteen representatives from twelve Member States and International Organizations attended the Meeting. This publication is a collection of presentations in which the participants reported the status of their national programmes on fast breeder reactors. A separate abstract was prepared for each of the twelve papers from this collections. Refs, figs, tabs and 1 graph

  5. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  6. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  7. Decay Heat Removal for the Liquid Metal Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. The paper closes with a statement of the high reliability of the Clinch River Breeder Reactor Plant decay heat removal systems and a summary of the supporting arguments. (author)

  8. Feasibility studies on commercialized fast breeder reactor cycle system

    International Nuclear Information System (INIS)

    JNC (Japan Nuclear Cycle Development Institute) and the electric utilities in Japan have established a new organization to develop a commercialized fast breeder reactor (FBR) cycle system since July 1, 1999, feasibility studies (F/S) have been undertaken in order to determine the promising concepts and to define the necessary R and D tasks. In the first two-year phase, a number of candidate concepts will be selected from various options, featuring innovative technologies. In the F/S, the options are evaluated and conceptual designs are examined considering the attainable perspectives for following: 1) ensuring safety, 2) economic competitiveness to future LWRs, 3) efficient utilization of resources, 4) reduction of environmental burden and 5) enhancement of nuclear non-proliferation. The F/S should also guide the necessary R and D to commercialize FBR cycle system. In particular enhanced technologies should be integrated in order to ensure nuclear non-proliferation. In the second five-year phase of the F/S, scaled engineering tests will be conducted. Based on the test data, a comprehensive evaluation will be conducted to confirm the technical attainability of candidate concepts. A few proposals for the commercialization of the FBR cycle system will be proposed. (author)

  9. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  10. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  11. Elements for evaluation of fast breeder reactor's potential in Argentina

    International Nuclear Information System (INIS)

    Fast Breeder Reactors (FBR) main features are presented in a general form, including their physical principles, the history of their evolution, their relevant technological aspects and the basis for their comparison to other energy sources. This is completed with descriptions of typical reactors and a model of FBR penetration in the Argentine electrical network. It is recommended to form a multidisciplinary board to study which position should be taken with respect to this type of reactors. In the author's opinion a Research activity should be started and gradually increased for passing to Development activities after a short while. (Author)

  12. Integral measurement of fission products capture in fast breeder reactors

    International Nuclear Information System (INIS)

    For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set

  13. Challenges and achievements - Prototype Fast Breeder Reactor construction

    International Nuclear Information System (INIS)

    Prototype fast breeder reactor presently under construction poses several challenges in materials, design and construction. The civil structure and equipment are of very large size and complex in nature. This paper presents the features of the design and construction of the PFBR excavation, raft, civil structure of the nuclear island connected buildings and reactor vault. This paper also brings out the details of the large size equipment of special stainless steel and handling structure for their lifting and placement inside the reactor vault. The paper is divided into three parts viz. introduction, challenges and achievements during construction of civil structures and erection of large size components. (author)

  14. Liquid metal fast breeder reactor: an environmental and economic critique

    International Nuclear Information System (INIS)

    Economic and environmental arguments made by the AEC and others for the liquid metal fast breeder reactor (LMFBR) as a central component of the U. S. electrical energy system are discussed. The LMFBR appears to have no environmental advantage over the currently operating light water reactor and especially not over the high temperature gas reactor. The principle environmental argument for the rapid introduction of LMFBRs is that they will provide a virtually inexhaustible fuel source, and reduce the demand for strip-mining the limited reserves of high grade U ore. A 20-yr delay in the construction of LMFBRs would result in an increase of only 50 mi2 of strip mining over the next 50 yr, and the cost of reclamation of this land would be about 0.1 mill/kw-hr. Uranium from which fuel has been extracted for use by nonbreeder reactors can still be used by breeders, thus breeders could still be introduced in the future, if fusion is not developed in time, and extract the same overall energy from a given supply of U as if they had been introduced earlier. Economic arguments in favor of the LMFBR are based on models highly sensitive to changes on some of the most critical input variables: nuclear power plant capital costs, fuel cycle costs, performance characteristics of LMFBR designs, electrical energy demand, and U ore costs. There is no basis for concluding that the LMFBR will be economical in the 1980s or early 1990s. (Pollut. Abstr.)

  15. Report to the Congress: liquid metal fast breeder reactor program--past, present, and future, Energy Research and Development Administration

    International Nuclear Information System (INIS)

    The past, present, and future of the liquid metal fast breeder reactor (LMFBR) program, the Nation's highest priority energy program, are studied. ERDA anticipates that the operation of the first large commercial breeder will start in 1987, and that 186 commercial-size breeders will be in operation by the year 2000. The breeder program is made up of six major areas, each dealing with an important element of technology: reactor physics; fuels and materials; fuel recycle; safety; component development; plant experience; and facilities used in the LMFBR program. ERDA is implementing a new system for administering, managing, and controlling the breeder program that will provide increased program visibility and control. Federal funding for breeder development was $168 million in FY 1971, accounting for 40% of the total Federal R and D energy budget; in FY 1976 Federal funding for this program will be $474 million, only 26% of total Federal funding for energy research. Besides Federal funds, over half a billion dollars have been or will be invested by industry over the next 5 to 10 years to develop the breeder and to build a demonstration plant. Five other nations--the United Kingdom, France, Japan, West Germany, and the Soviet Union--have a high priority national energy program for developing the LMFBR. These foreign breeder programs could contribute important data and information to the U.S. program

  16. The fast breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    This paper outlines the current national fast reactor program in France and U.K and describes the increasing plant operational experience being acquired in the two countries for fuel reprocessing and the European project of a series of demonstration reprocessing plants of sufficient capacity to serve the needs of several commercially sized fast reactors. The key futures of France and U.K. programs are: fuel dismantling and pin cropping, dissolution, fuel dissolvers, liquor clarification, plutonium accountancy, solvent extraction, product preparation and packaging, wastes and emissions and fuel fabrication (initial blending, milling, pellet pressing, etc...)

  17. Fast breeder reactors: Experience and trends. V. 2

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

  18. Liquid Metal Fast Breeder Reactor program. Volume III. Environmental statement

    International Nuclear Information System (INIS)

    The various alternative technologies, nuclear as well as nonnuclear, that might be utilized in conjunction with or instead of the LMFBR to satisfy the Nation's future electric power requirements are examined. The options considered include the further implementation of various types of nuclear power reactors such as the already existing light water reactor and high temperature gas-cooled reactor, as well as the development of alternative breeder reactors such as the gas-cooled fast reactor, light water breeder reactor and molten salt breeder reactor. The development of another potential nuclear energy system, controlled thermonuclear fusion, is also addressed. The possibilities of increased emphasis on the use of conventional fossil fuels, namely coal, oil and natural gas, and the development of unconventional fossil fuels such as oil shale and domestic tar sands are discussed, followed by consideration of the further development of additional nonnuclear energy sources such as hydroelectric power systems, geothermal energy, solar energy, and other potential sources of power. Each option is examined as to the extent of its energy resource base, the research and development program that would be required (if any) to bring the option into commercial use, the environmental implications of its utilization and the costs and benefits associated with its use, in order to assess its capability for satisfying projected energy requirements. The use of improved energy conversion and storage devices such as gas turbines, fuel cells and magnetohydrodynamics is discussed. An examination of the various elements of a potential national effort in energy conservation to assess their capabilities for reducing projected energy demands and thereby replacing partially or entirely the need for additional power sources such as the LMFBR is presented. (U.S.)

  19. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1989 as reported at the 23rd meeting of the IWGFR in Vienna, April 1990. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States. A separate abstract was prepared for each of the 11 papers presented by the participants of this meeting. Refs, figs and tabs

  20. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1990 as reported at the 24th meeting of the IWGFR in Tsuruga, Japan, 15-18 April 1991. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States and CEC. Figs and tabs

  1. Thermal and neutronic calculation for fast breeder reactor FBR

    International Nuclear Information System (INIS)

    This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps

  2. Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

  3. Plutonium breeding in liquid-metal fast breeder reactors and light water reactors

    International Nuclear Information System (INIS)

    The possibilities of breeding in liquid-metal fast breeder reactors (LMFBRs) and light water reactors (LWRs) are compared in two ways. The feasibility of breeding has been demonstrated in the Phenix reactor with a measured gain of 0.14. The gain in Superphenix will amount to about0.20. The studies show that while maintaining the performance of commercial reactors their breeding gain can be further increased either by the concept of heterogeneous cores or by using carbide or nitride fuel (breeding gain about0.35). Recently, the old idea of breeding in advanced pressurized water reactors (PWRs) has been taken up again with the objective of attaining a gain of 0.05. Unfortunately, these objectives had to be limited to a conversion ratio of 0.9 for safety reasons, and it is not certain whether operation will be rewarding economically. The strategy of substituting PWRs is examined using the French example. By gradually introducing LMFBRs, the cumulated uranium supplies in France can be kept within reasonable limits, which means that they attain three to four times the home resources. This is not possible with advanced LWRs, which can be considered only as a possible backup solution for plutonium recycling into PWRs

  4. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2009-05-01

    In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.

  5. Neutron irradiation of candidate ceramic breeder materials of fusion reactors

    International Nuclear Information System (INIS)

    In the context of the European programs for the future fusion reactors, the Process Chemistry Department of ENEA, Casaccia Center (Rome), has been involved in preparing ceramic blanket materials as tritium breeders; a special consideration has been addressed to the nuclear characterization of LiAlO2 and Li2ZrO3. In this paper are reported neutron irradiation of ceramic specimens in TRIGA reactor and γ-spectrometric measurements for INAA purposes; and isothermal annealing of the irradiated samples and tritium extraction, by using an 'out of pile' system. (author) 3 refs.; 4 figs.; 4 tabs

  6. Reliability modeling of Clinch River breeder reactor electrical shutdown systems

    International Nuclear Information System (INIS)

    The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

  7. Sodium technology for fast breeder reactors

    International Nuclear Information System (INIS)

    Sodium, because of its good heat transfer and nuclear properties, is used as a coolant in fast reactors. It is also used largely as a reducing agent in pharmaceutical, perfumery and general chemical industries. Its affinity to react with air and water is a strong disadvantage. However, this is fully understood and the design of engineering systems take care of this aspect. With several experimental and test facilities established over the years in this country as well as abroad, the 'sodium technology' has reached a level of maturity. The design of sodium systems considering all the physical and chemical properties and the developmental work carried out at Indira Gandhi Centre for Atomic Research are broadly covered in this report. (author)

  8. Feasibility and deployment strategy of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    The author have studied water cooled thorium breeder reactor based on matured pressurized water reactor (PWR) plant technology for several years. Through these studies it is concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array with heavy water coolant in the primary loop by replacing original light water is appropriate for achieving sufficient breeding performance as sustainable fission system and high enough burn-up as an economical power plant. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste

  9. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6/ ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

  10. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

    A reactor system for producing useful thermal energy and valuable isotopes, such as plutonium-239, uranium-233, and/or tritium, in which a pair of sub-critical masses of fissile and fertile actinide slugs are propelled into an ellipsoidal pressure vessel. The propelled slugs intercept near the center of the chamber where the concurring slugs become a more than prompt configuration thereby producing a fission explosion. Re-useable accelerating mechanisms are provided external of the vessel for propelling the slugs at predetermined time intervals into the vessel. A working fluid of lean molten metal slurry is injected into the chamber prior to each explosion for the attenuation of the explosion's effects, for the protection of the chamber's walls, and for the absorbtion of thermal energy and debris from the explosion. The working fluid is injected into the chamber in a pattern so as not to interfere with the flight paths of the slugs and to maximize the concentration of working fluid near the chamber's center. The heated working fluid is drained from the vessel and is used to perform useful work. Most of the debris from the explosion is collected as precipitate and is used for the manufacture of new slugs

  11. Plutonium bearing oxide fuels for recycling in thermal reactors and fast breeder reactors

    International Nuclear Information System (INIS)

    Programs carried out in the past two decades have established the technical feasibility of using plutonium as a fuel material in both water-cooled power reactors and sodium-cooled fast breeder reactors. The problem facing the technical community is basically one of demonstrating plutonium fuel recycle under strict conditions of public safety, accountability, personnel exposure, waste management, transportation and diversion or theft which are still evolving. In this paper only technical and economic aspects of high volume production and the demonstration program required are discussed. This paper discusses the role of mixed oxide fuels in light water reactors and the objectives of the LMFBR required for continual growth of nuclear power during the next century. The results of studies showing the impact of using plutonium on uranium requirements, power costs, and the market share of nuclear power are presented. The influence of doubling time and the introduction date of LMFBRs on the benefits to be derived by its commercial use are discussed. Advanced fuel development programs scoped to meet future commerical LMFBR fuel requirements are described. Programs designed to provide the basic technology required for using plutonium fuels in a manner which will satisfy all requirements for public acceptance are described. Included are the high exposure plutonium fabrication development program centered around the High Performance Fuels Laboratory being built at the Hanford Engineering Development Laboratory and the program to confirm the technology required for the production of mixed oxide fuels for light water reactors which is being coordinated by Savannah River Laboratories

  12. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Baldev Raj; S L Mannan; P R Vasudeva Rao; M D Mathew

    2002-10-01

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.

  13. Experiences with fast breeder reactor education in laboratory and short course settings

    International Nuclear Information System (INIS)

    The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented

  14. Computational intelligent systems for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Nearly 15000 process signals are digitized by physically and functionally distributed embedded systems in Prototype Fast Breeder Reactor (PFBR). Digitized signals are processed and relevant information is displayed through Large video display systems at Control Room. It is necessary that correct and reliable information need to be provided to the plant operator. Computational intelligent systems play a major role in enhancing the safe operation of the Nuclear reactor. The paper explains the features of three such systems, one for on-line validation of neutronic power channel through on-line thermal balance calculation and another for detection of anomalous reactivity addition through on-line reactivity balance computation and third for on-line computation of Reactor power from fluctuations of core thermocouple signals. (author)

  15. Fabrication of metallic fuel for fast breeder reactor

    International Nuclear Information System (INIS)

    Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

  16. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  17. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749

  18. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    Problems of public acceptance of nuclear power have affected the development of fast breeder reactors in the Federal Republic of Germany. Besides cooperation with UK, USA and Japan, the most outstanding event in the field of international fast breeder cooperation was a set of the agreements between Germany and France. These agreements opened the possibility of joint fast breeder development by Germany together with Belgium and the Netherlands. Most activities on the site of Compact Sodium Cooled Nuclear Reactor KNK-II were concerned with commissioning of the plant and final construction work. Criticality was achieved in Oct. 1977 and low-power tests performed. This paper includes a description of the status of construction of SNR-300 reactor and the results of research and development programmes performed. These were concerned with fuel elements development and results of irradiation experiments; development of cladding materials and core element structural materials; interaction between fuel and cladding; sodium tests; development and verification of computer codes; experiments in fast critical assemblies; fast rector safety; core disruptive accidents; development of instrumentation; thermodynamics od fuel assemblies; fluid dynamics

  19. Liquid metal cooled fast breeder nuclear reactor constructions

    International Nuclear Information System (INIS)

    A description is given of a liquid metal cooled fast breeder nuclear reactor construction of the pool kind in which the primary vessel incorporates an annular yoke fabricated from arcuate segments. The yoke is suspended from the roof structure of the vault by a first annular series of tie straps arranged outside the primary vessel whilst a strongback on which the fuel assembly sits inside the primary vessel is supported from the yoke by a second series of tie straps. The yoke has upwardly and downwardly extending legs which are extended by upper and lower strakes respectively of the primary vessel. (U.K.)

  20. Experience with EBR-II [Experimental Breeder Reactor] driver fuel

    International Nuclear Information System (INIS)

    The exceptional performance of Experimental Breeder Reactor-II (EBR-II) metallic driver fuel has been demonstrated by the irradiation of a large number of elements under steady-state, transient overpower, and loss-of-flow conditions. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. Quantification of reliability has allowed full utilization of element lifetime. Improved design and duct materials currently under test are expected to increase the burnup from 8 to 14 at.%

  1. Blanket management method for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    A method for reducing thermal striping in liquid metal fast breeder reactors by reducing temperature gradients between adjacent fuel and blanket assemblies by shuffling blanket assemblies at each refueling outage so as to progressively shuffle the blanket assemblies to the core periphery through multiple moves and to generally locate fresh blanket assemblies adjacent to exposed fuel assemblies and exposed blanket assemblies adjacent to fresh fuel. Additionally, assembly orificing is altered to provide less flow to blanket assemblies needing less flow due to an otherwise decreased temperature gradient and providing additional flow to fuel assemblies which need more flow to sufficiently reduce temperature gradients to prevent thermal striping. (author)

  2. Clinch River Breeder Reactor secondary control rod system

    International Nuclear Information System (INIS)

    The shutdown system for the Clinch River Breeder Reactor (CRBR) includes two independent systems--a primary and a secondary system. The Secondary Control Rod System (SCRS) is a new design which is being developed by General Electric to be independent from the primary system in order to improve overall shutdown reliability by eliminating potential common-mode failures. The paper describes the status of the SCRS design and fabrication and testing activities. Design verification testing on the component level is largely complete. These component tests are covered with emphasis on design impact results. A prototype unit has been manufactured and system level tests in sodium have been initiated

  3. Base isolation system for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    The use of seismic isolation specially in the high seismic regions has gained increasing interest as a viable and efficient solution to earthquake ground motion both within and outside the nuclear field. A feasibility study to see the effect of laminated rubber bearing pads for the 500 MWe pool type fast breeder reactor has been carried out. The results show that there is 2 to 2.5 times reduction in floor response spectra peak and the seismic loads on the components are considerably reduced. The problem areas include the potential for the large sloshing amplitudes, accommodating large displacements in the piping etc. (author)

  4. Operating and test experience with Experimental Breeder Reactor number 2 (EBR-II), the Integral Fast Reactor (IFR) prototype

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor number 2 (EBR-II) has operated for 30 years, the longest for any liquid metal cooled reactor (LMR) power plant in the world. Given the scope of what has been developed and demonstrated over those years, it is arguably the most successful test reactor operation ever. Tests have been carried out on virtually every fast reactor fuel type. The reactor itself has been extensively studied. The most dramatic safety tests, conducted on 3 April, 1986, showed that an LMR with metallic fuel could safely accommodate loss of flow or loss of heat-sink without scram. EBR-II operated as the Integral Fast Reactor (IFR) prototype, demonstrating important innovations in safety, plant design, fuel design and actinide recycle. The ability to accommodate anticipated transients without scram passively resulted in significant simplification of the reactor plant, primarily through less reliance on emergency power and not having to require the secondary sodium or steam systems to be safety grade. These features have been quantified in a probabilistic risk assessment (PRA) conducted for EBR-II, demonstrating considerable safety advantages over other reactor concepts. Fundamental to the superior safety and operating characteristics of this reactor is the metallic U-Pu-Zr alloy fuel. Performance of the fuel has been fully proven: achieved burnup levels exceed 20 at.% in the lead test assemblies. A complete set of fuel performance and safety limits has been developed and was carried forward in formal safety documents supporting conversion of the core to IFR fuel. The last major demonstration planned was to assess the performance of recycled actinides in the fuel and to confirm that passive safety characteristics are maintained with recycled actinide fuel in the core. (author)

  5. It is now time to proceed with a gas-cooled breeder reactor (GBR) demonstration plant

    International Nuclear Information System (INIS)

    Since 1969, the GBRA has been making studies to provide evidence on questions which were not clear regarding the Gas-cooled Breeder Reactor: design feasibility and performance, safety, strategy and economics, and R and D necessary for a demonstration plant. Studies were carried out on a 1200-MW(e) commercial reference design with pin fuel, which was also used as a basis for a definition of the GBR demonstration plant. During the six years, a great deal of information has been generated at GBRA and it confirms the forecasts of the promoters of the Gas-cooled Breeder Reactor that the GBR is an excellent reactor from all points of view: design - the reactor can be engineered without major difficulty, using present techniques. As far as fuel is concerned, LMFBR fuel technology is applicable plus limited specific development effort. Performance - the GBR is the best breeder with oxide fuel and using conventional techniques. The strategy studies carried out at GBRA have clearly shown that a high performance breeder such as the GBR is absolutely required in large quantities by the turn of the century in order to avoid dependence on natural uranium resources. Regarding safety, a major step forward has been made when an ad hoc group on GBR safety, sponsored by the EEC, concluded that no major difficulties were anticipated which would prevent the GBR reaching adequate safety standards. Detailed economic assessments performed on an item-to-item basis have shown that the cost of a GBR with its high safety standard is about the same as that of an HTR. One can therefore conclude that, with the present cost of natural uranium, the GBR is competitive with the LWRs. Owing to the very limited R and D effort necessary and the obvious safety, economic and strategic advantages of the concept, it is concluded that the development and construction of a GBR demonstration plant must be started now if one wants to secure an adequate energy supply past the turn of the century. (author)

  6. Fast Breeder Test Reactor: 15 years of operating experience

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 1985 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 1993 and power was raised to 10.5 MWt in Dec 1993. Turbine generator was synchronized to the grid in Jul 1997. The indigenously developed mixed carbide fuel has achieved a peak burn up of 88,000 MWd/t till now at a linear heat rating of 320 W/cm and reactor power of 13.4 MWt without any fuel-clad failure. The paper presents operating and decontamination experience, performance of fuel, steam generator and sodium circuits, certain unusual occurrences encountered by the plant and various improvements carried out in reactor systems to enhance plant availability. (author)

  7. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    P K Vijayan; A Basak; I V Dulera; K K Vaze; S Basu; R K Sinha

    2015-09-01

    The third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th-based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.

  8. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  9. Fast breeder reactors: experience and trends. V. 1

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium presentations were divided into sessions devoted to the following topics: Experience of LMFBR construction and operation and resultant development strategies (6 papers); LMFBR plant startup and commissioning tests and general behaviour (8 papers); Core performance experience for high burnup and core design trends (8 papers); Experience and trends in the LMFBR fuel cycle (4 papers); Core design and behaviour (3 papers); Fuels and materials (7 papers). A separate abstract was prepared for each of these papers

  10. Breeder reactors: a technique at the service of humanity

    International Nuclear Information System (INIS)

    A genuine energy policy is not conceived purely for a short term. It must on the contrary take into consideration many national and international facts in order to arrive at a balance which takes into account both the interests of the country where it is to be applied and the future interests of humanity. Growth and energy consumption make a pair. Considering the forecasts of future consumption, a rational utilization of the energy sources is a priority. The rational utilization of the energy potentialities of uranium takes a prominent place in this priority. In the fission energy of the atoms, the breeder reactors are the only types which can give their full meanings to the words economy, ecology, rationality etc. In calling for innovation, the breeder reactors are the prime movers for an advanced industry and a guarantee for the future penetration of electricity in many fields. They are thus important elements for the creation of employment. This paper also deals with questions of international cooperation, non-proliferation and the necessity for disarmament

  11. Characterizing the tribological behaviour of fast breeder reactor materials

    International Nuclear Information System (INIS)

    The object of these tests is to define the behaviour of material couples working in conditions as representative as possible of reactor operation. For this purpose a certain number of test installations have been developed to simulate the most typical cases of friction encountered: plane to plane geometry, rotational bearings, guiding bearings. Endurance tests have also been carried out on ball bearings and ballscrews samples. As said before, the test conditions attempt to reproduce as faithfully as possible the environment of the materials used in fast breeder reactors, particularly in: - using purified liquid sodium, and maintaining it isotherm, respectively at three temperature levels: 180, 400 and 5500C; - or using argon containing sodium aerosol particles. Some typical values of friction coefficients and rates of wear obtained during the tests with certain couples of materials are given here as examples. The aims which are currently guiding the direction of the tests are also briefly described

  12. A linear model of the Fast Breeder Test Reactor Plant

    International Nuclear Information System (INIS)

    A linear analysis of the Fast Breeder Test Reactor System, consisting of the reactor, intermediate heat exchanger, steam generator and connected piping is presented. The problem of variable boundaries in the steam generator is reduced to a problem of fixed boundaries by dividing the steam generator into six zones. Based upon this, one can obtain the transfer function of any input/output combination. Starting with the time domain non-linear partial differential equations, the problem is reduced to a system of linear equations in complex variables, which can be solved basically by Gaussian elimination process. The results of this work will be useful in determining a suitable control scheme for waterflow in the steam generator and the control parameters. (auth.)

  13. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  14. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Fast Breeder Reactor-Block Antiseismic Design and Verification'' was organized by the ENEA Fast Reactor Department in co-operation with the International Working Group (IWGFR) of the International Atomic Energy Agency (IAEA), according to the recommendations of the 19th IAEA/IWGFR Meeting. It was held in Bologna, at the Headquarters of the ENEA Fast Reactor Department, on October 12-15, 1987, in the framework of the Celebrations for the Ninth Centenary of the Bologna University. The proceedings of the meeting consists of three parts. Part 1 contains the introduction and general comments, the agenda of the meeting, session summaries, conclusions and recommendations and the list of participants. Part 2 contains 8 status reports of Member States participating in the Working Group. Contributed papers were published in Part 3 and were further subdivided into 5 sessions as follows: whole reactor-block analysis (4 papers); whole reactor-block analysis (sloshing and buckling, seismic isolation effects) (8 papers); detailed core analysis (6 papers); shutdown systems and core structural and functional verifications (6 papers); component and piping analysis (7 papers). A separate abstract was prepared for each of the 8 status reports and 31 contributed papers. Refs, figs and tabs

  15. Experience with the generating plant at fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWth/13.2 MW(e) sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and generating systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct.1985 with Mark-I core (70 % PUC-30 % UC). FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. Heat generated in the reactor core is transported to the tertiary loop by primary and secondary sodium loops. The steam water system mainly consists of a once through steam generator, which produces super heated steam at a pressure of 120 bars and temperature of 480 degC, feed water system and condensate system. The steam produced is supplied to a condensing turbine. The turbine in turn is coupled to an alternator. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. The paper presents operating experience with generating plant consisting of steam water circuit, condensing turbine and its associated systems and the alternator, various modifications carried out to improve system reliability and availability and certain incidents taken place in the generating plant. (author)

  16. Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

    International Nuclear Information System (INIS)

    The seismic analysis of reactor assembly housing the primary circuit of a typical 500 MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor assembly by detailed finite element analysis including the fluid-structure interaction and sloshing effects. Analysis is carried out to quantify the effects of inter-connection of the reactor vault with the adjacent buildings under the assumptions that the reactor vault along with reactor assembly is: (1) an isolated structural system from the adjacent buildings within reactor containment building (RCB) and (2) connected with the adjacent civil structures through floor slabs. Analysis indicates that, by inter-connecting the vault with the NICB, there are overall increases of all the governing parameters which decide the seismic design criteria. The significant effects are increases of: (1) radial and axial displacements of core top and absorber rods and vertical accelerations of core subassemblies which are of concern to reactor safety, (2) primary membrane stress intensities for the inner vessel and (3) strain energies developed at the critical portions which can enhance the buckling risks of main vessel, inner vessel and thermal baffles. Hence, it is preferable to isolate the reactor vault, directly constructing from the base raft without inter-connecting it with the NICB, from the seismic loading considerations

  17. Immediate relation of ING to fast breeder reactor programs

    International Nuclear Information System (INIS)

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  18. Design of Central Sub Assembly Temperature Monitoring System for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The Central Sub Assembly Temperature Monitoring (CSATM) System for 500 MWe Prototype Fast Breeder Reactor (PFBR) is a safety critical system. It is an independent, standalone, hardwired and diversified system to neutronic parameters. The detection of integrity of the subassembly plays a major role, because of high power density and compact core structure of PFBR fuel. To achieve this, CSATM system is provided for the measurement and detection of overshoot for Central Sub Assembly temperature. It protects the reactor from various incidents such as transient overpower at low power and high power, blockage of coolant, pipe rupture etc. CSATM system with triple modular redundancy is employed to measure the central sub-assembly outlet temperature (θCSA) and safety action will be initiated if temperature reaches beyond SCRAM threshold level. (author)

  19. Shutdown and Closure of the Experimental Breeder Reactor - II

    International Nuclear Information System (INIS)

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor - II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated lay-up plan defining the system end state, as well as instructions for achieving the lay-up condition. A goal of system-by-system lay-up is to minimize surveillance and

  20. Shutdown and closure of the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and

  1. Implementation of multivariable control techniques with application to Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    After several successful applications to aerospace industry, the modern control theory methods have recently attracted many control engineers from other engineering disciplines. For advanced nuclear reactors, the modern control theory may provide major advantages in safety, availability, and economic aspects. This report is intended to illustrate the feasibility of applying the linear quadratic Gaussian (LQG) compensator in nuclear reactor applications. The LQG design is compared with the existing classical control schemes. Both approaches are tested using the Experimental Breeder Reactor 2 (EBR-2) as the system. The experiments are performed using a mathematical model of the EBR-2 plant. Despite the fact that the controller and plant models do not include all known physical constraints, the results are encouraging. This preliminary study provides an informative, introductory picture for future considerations of using modern control theory methods in nuclear industry. 10 refs., 25 figs

  2. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  3. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  4. Accident analysis of heavy water cooled thorium breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  5. Accident analysis of heavy water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  6. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value

  7. Development of sodium facilities for NSRR fast breeder reactor fuel tests. 2. Sodium capsule

    International Nuclear Information System (INIS)

    In order to commercialize fast reactors, which are expected to be long-term transmutes of plutonium and long half life radioactive wastes (such as americium) from light water reactors, safety research under accident conditions and establishment of the safety guidelines are essential. Sodium facilities, such as, (1) Purification/charging loop and test loop, and (2) Proto-type Sodium capsule, were developed and fabricated in order to pulse irradiate fast breeder reactor fuels in the Nuclear Safety Research Reactor (NSRR) of JAERI for investigation on fuel behavior under transient over-power conditions. This report presents the purpose, outlines, specifications, capabilities and operation results of the proto-type sodium capsule. Two kinds of capsule, i.e., the stagnant sodium capsule and the sodium loop, were designed to pulse irradiate Fast Reactor (FR) fuels in the NSRR under sodium cooling conditions with and without flow, respectively. Because the capsules have to safely contain chemically active sodium at high temperature and stand the pressure pulses by the sodium hummer which might be generated at fuel failure, the development of the capsule is essential for realizing the research. Thus, proto-type sodium loop, which consisted of doubly sealed container, sodium pump and flow meter, was developed. In addition, two type of flange structure for the stagnant capsule and loop was leak tested at high pressures, in order to confirm its sealing capability at room and high temperature conditions. (author)

  8. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  9. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  10. FPGA based pump speed measurement system for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    In Prototype Fast Breeder Reactor (PFBR), the heat generated in the reactor core by nuclear fission is extracted by circulating liquid sodium through it using two Primary Sodium Pumps (PSP). The FPGA-based PSP Speed measurement system is a safety critical system provided to protect the reactor in the event of a PSP seizure. The function of the system is to measure the PSP speed and initiate safety action if there is a reduction in speed below a specified set point. Variable Reluctance Sensor (VRS) is used for measuring the PSP speed. This sensor outputs a voltage signal, whose amplitude and frequency are proportional to the pump speed. The frequency of the sensor signal is measured, translated to the pump speed, and compared with user-specified set points for generating the required alarm and trip (safety action) signals. This paper explains the system requirements, system architecture, implementation and qualification tests carried out on the system. Since the timing requirements on the system are stringent, a pipelined architecture is used for improving the system response time, which is detailed in this paper. Since the system is safety critical, various safety and failsafe features are incorporated in the system which are also explained. (author)

  11. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  12. Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

  13. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    International Nuclear Information System (INIS)

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR

  14. Flow induced vibrations in liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Flow induced vibrations are well known phenomena in industry. Engineers have to estimate their destructive effects on structures. In the nuclear industry, flow induced vibrations are assessed early in the design process, and the results are incorporated in the design procedures. In many cases, model testing is used to supplement the design process to ensure that detrimental behaviour due to flow induced vibrations will not occur in the component in question. While these procedures attempt to minimize the probability of adverse performance of the various components, there is a problem in the extrapolation of analytical design techniques and/or model testing to actual plant operation. Therefore, sodium tests or vibrational measurements of components in the reactor system are used to provide additional assurance. This report is a general survey of experimental and calculational methods in this area of structural mechanics. The report is addressed to specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors. 92 refs, 90 figs, 8 tabs

  15. Experience of secondary cooling system modification at prototype fast breeder reactor MONJU (Translated document)

    International Nuclear Information System (INIS)

    The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident that occurred in December 1995. After the accident, an investigation into the cause and a comprehensive safety review of the plant were conducted, and various countermeasures for sodium leak were examined. Modification work commenced in September 2005. Since sodium, a chemically active material, is used as coolant in MONJU, the modification work required work methods suitable for the handling of sodium. From this perspective, the use of a plastic bag when opening the sodium boundary, oxygen concentration control in a plastic bag, slightly-positive pressure control of cover gas in the systems, pressing and cutting with a roller cutter to prevent the incorporation of metal fillings, etc. were adopted, with careful consideration given to experience and findings from previous modification work at the experimental fast reactor JOYO and plants abroad. Owing to these work methods, the modification work proceeded close to schedule without incident. (author)

  16. Network representation of design knowledge of prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    A method of design knowledge representation was studied for the Japanese fast breeder reactor Monju, aiming at enhanced understanding of engineering considerations with mutual relations. Taking over design knowledge of Monju to next generation designers/engineers to be in charge of design of future FRs is by no means easy, in contrast with operation and maintenance knowledge which can be acquired in the real plant operation and maintenance. Specifications of the as-is Monju contains only a small part of the entire design knowledge, mainly by two reasons. Firstly, reasons for selecting the as-is specifications can not be understood until reaching proper knowledge source. Secondly, there are many passed-over options on the design specifications. Reasons for passing-over these options are not always technical inferiority. A large part of the current specifications are selected because the worst possible technical value can be foreseeable or guaranteed to be acceptable within limited R and D period and resource, not because the expected value is estimated to be the lower. In other words, in the future where new materials with improved properties, faster and more accurate analysis/prediction methods, rationalized technical standards or regulatory requirements, and/or some other environment for thorough comparison among specification options are available, these passed-over options are likely to be worth reconsidering. There are a huge number of technical documents on diversified engineering studies, such as calculation of maximum possible temperature gradient of important structures, necessary sodium flow rate in particular sub-assemblies, etc. for validation of each decision making in design. A large part of these documents are scanned and stored in a data base with each catalogue data for electronic browse. The authors propose a network representation of these items of design decision making, where the items are mutually connected by directed arcs, where nodes stand

  17. Reactor Safety: Introduction

    International Nuclear Information System (INIS)

    The programme of the Reactor Safety Division focuses on the development of expertise on materials behaviour under irradiation for fission and fusion oriented applications. Furthermore, as nuclear energy needs international public acceptance with respect to safety and efficient management of natural resources and wants to reduce the burden of nuclear waste, the Reactor Safety Division enhanced its efforts to develop the MYRRHA project. MYRRHA, an accelerator driven sub-critical system, might have the potential to cope in Europe with the above mentioned constraints on acceptability and might serve as a technological platform for GEN IV reactor development, in particular the Liquid Metal Fast Reactor.The Reactor Safety Division gathers three research entities that are internationally recognised: the Reactor Materials Research department, the Reactor Physics and MYRRHA department and the Instrumentation department.The objectives of Reactor Materials Research are: to evaluate the integrity and behaviour of structural materials and nuclear fuels used in present and future nuclear power industry; to perform research to unravel and understand the parameters that determine the material and fuel behaviour under or after irradiation; to contribute to the interpretation and modelling of the materials and fuels behaviour in order to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the Reactor Materials Research department concentrate on four distinct disciplines: Reactor Pressure Vessel Steel embrittlement Stress corrosion cracking in reactor coolant environment, including Irradiation Assisted Stress Corrosion Cracking; Nuclear Fuel characterisation and development of new fuel types for commercial and test reactors. Development of materials for Fusion and advanced nuclear fission reactors. The safe operation of present nuclear power plants relies primarily on the integrity of the reactor pressure vessel

  18. Design optimization of backup seal for sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: ► Design arrived from fourteen geometric options by finite element analysis. ► Seal geometry, size, compression, contact pressure, stress and compression load optimized. ► Effects of reduced fluoroelastomer strength at 110 °C, strain rate and stress-softening incorporated. ► Ageing, friction, tolerances, batch-to-batch/production variations in fluoroelastomer considered. ► Procedure applicable to other elastomeric seals of Fast Breeder Reactors. -- Abstract: Design optimization of static, fluoroelastomer backup seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) is depicted. 14 geometric variations of a solid trapezoidal cross-section were studied by finite element analysis (FEA) to arrive at a design with hollowness and double o-ring contours on the sealing face. The seal design with squeeze of 5 mm assures failsafe operation for at least 10 years under a differential pressure of 25 kPa and ageing influences of fluid (air), temperature (110 °C) and γ radiation (23 mGy/h) in reactor. Hybrid elements of 1 mm length, regular integration, Mooney–Rivlin material model and Poisson’s ratio of 0.493 were used in axisymmetric analysis scheme. Possible effects of reduced fluoroelastomer strength at 110 °C, ageing, friction, tolerances in reactor scale, testing conditions during FEA data generation and batch-to-batch/production variations in seal material were considered to ensure adequate safety margin at the end of design life. The safety margin and numerical prediction accuracy could be improved further by using properties of specimens extracted from seal. The approach is applicable to other low pressure, moderate temperature elastomeric sealing applications of PFBR, mostly operating under maximum strain of 50%.

  19. Manufacturing of prototype fast breeder reactor components: challenges and achievements

    International Nuclear Information System (INIS)

    In the presentation, three components of 500 MWe Prototype Fast Breeder Reactor (PFBR), viz. grid plate, roof slab and fuel handling systems, are focused, which have been responsible for the considerable delay of the project schedule. The manufacturing challenges of grid plate mainly originated from large number of sleeves resulting in higher self weight and hard facing of large diameter sleeves. Machining of large diameter plates and shell assembly to the required tight tolerances on dimensions, hard facing with nickel based cobalt free hard facing material on continuous, large diameter (6.7 m) annular tracks, heat treatment of large austenitic stainless steel parts at 1050℃ with controlled rates of cooling and heating together with control on temperature gradient across the parts, complex assembly of a large number of parts (∼14900) meeting the important requirements on verticality of sleeve assemblies (Ø0.1 mm) and delicate handling and transportation are truly challenging activities in the manufacturing technology. In case of roof slab, complex manufacturing process, especially welding between the shell and stiffeners caused lamellar tearing problems and extensive testing time. Inclined fuel transfer machine, multiple repairs, heavy weight and testing strategy resulted in long manufacturing and testing time. Some general lessons learnt are also brought out in this presentation. Technology development prior to start of construction is essential for long delivery components. Judicious choice of tolerances, number and location of welds and inspections has to be made. Robust criteria need to be applied for the acceptance of manufacturing deviations and material compositions. Indigenous materials should be used after qualifications of manufacturing process of direct relevance apart from routine standards. From the rich experience gained through the manufacture and erection of reactor assembly components of PFBR, important guidelines and approaches were derived

  20. Health physics experiences in the operation of Fast Breeder Test Reactor (FBTR)

    International Nuclear Information System (INIS)

    This paper presents the health physics experience gained with the operation of the Fast Breeder Test Reactor (FBTR), which was made critical in October 1985. Major operations that were carried out and the associated health physics surveillance are highlighted. (author)

  1. Reflections on the introduction of fast breeder reactors in the DeBeNeLux states

    International Nuclear Information System (INIS)

    This report gives a survey of the impact of introducing sodium-cooled fast breeder reactors in the Federal Republic of Germany and the BeNeLux countries (DeBeNeLux region). The supply situation with respect to electric and thermal energy is studied in particular, together with aspects of economy and environmental impact. The potential and consequences of a breeder economy, the present status and future r+d work are discussed. In addition to sodium-cooled fast breeder reactors with oxide or carbide fuel, alternative solutions are touched: 1) light water and high temperature reactors, 2) helium-cooled fast breeder reactors, 3) geothermal energy, solar energy and fusion energy. (orig.)

  2. Reactor system safety assurance

    International Nuclear Information System (INIS)

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  3. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    As for the experimental fast reactor ''Joyo'', the power increase test has been carried out since April, and the power output was raised stepwise up to 40 MW. The power output, core behavior, plant characteristics as well as shielding integrity were measured at each power level. The examination for licensing the power increase to 75 and 100 MW is still continued by the Committee No. 130. The preparation of various codes required for the core characteristic analysis is in progress. As for the development of the prototype fast reactor ''Monju'', the Construction Preliminary Design (1) was evaluated, and the studies on the specifications of the Construction Preliminary Design (2) are carried out. In respect to the analysis for the Safety Licensing, the analysis of decay heat, the development of an analytical code regarding the rupture propagation in heat transfer tubes for steam generators and others are under way. Technological investigation is carried out to obtain the overseas informations on the safety standards for FBRs and LMFBR technologies. The technical specifications for the preliminary design of the demonstration fast reactor are being prepared. The researches and developments of reactor physics, the structural components of ''Joyo'' and ''monju'', instrumentation and control, sodium technology, fuel materials, structural materials, safety and steam generators are reported, respectively. (Kako, I.)

  4. Development of Genetic Algorithm based Neural Network model for parameter estimation of Fast Breeder Reactor Subsystem

    OpenAIRE

    Subhra Rani Patra; R. Jehadeesan; Rajeswari, S.

    2012-01-01

    This work provides the construction of Genetic Algorithm based Neural Network for parameter estimation of Fast Breeder Test Reactor (FBTR) Subsystem. The parameter estimated here is temperature of Intermediate Heat Exchanger of Fast Breeder Test Reactor. Genetic Algorithm based Neural Network is a global search algorithm having less probability of being trapped in local minimum problem as compared to Standard Back Propagation algorithm which is a local search algorithm. The various developmen...

  5. Dynamic modeling of steam water system of prototype fast breeder reactor using RELAP code

    International Nuclear Information System (INIS)

    Highlights: • Dynamic modeling of steam water system of an LMBFR using RELAP5 code. • Analysis of events managed by power setback procedure in the plant. • Selection of parameters and their thresholds for the power setback procedure. - Abstract: The safety of a fast breeder reactor based nuclear power plant with an intermediate coolant loop between the primary and tertiary circuit, depends on the correct functioning of actions initiated by the balance-of-plant (tertiary) systems. Comfortable time will be available for such actions to ensure reactor safety. Perturbations in the Balance-of-Plant (BoP) influence the transient sequence of safety-relevant parameters of the plant in a benign manner. However, for complete and realistic prediction of transient behavior of the whole plant, dynamic models for BoP systems are required to be developed. This paper describes modeling of BoP system of PFBR using RELAP5/MOD 3.4 code. Some of the important transients in the BoP system, which are managed by reactor power setback procedure have been analysed using this code to verify the effectiveness of the procedure adopted

  6. Development of chemical sensors for Fast Breeder Reactor Technology

    International Nuclear Information System (INIS)

    Fast breeder reactors use liquid sodium as heat transfer medium and generate high pressure steam at the steam generator to run the turbine. This high pressure steam is separated from sodium coolant by ferritic steel tubes of 4 to 5 mm wall thickness. Development of any material defect in these heat exchanger tubes during their service would result in the ingress of high pressure steam into the sodium circuit leading to sodium-water reactions. A high temperature electrochemical hydrogen sensor based on CaBr2-CaHBr solid electrolyte and capable of measuring ppb levels of dissolved hydrogen in sodium has been developed at the laboratory. A very sensitive system, using thermal conductivity detector and semiconducting oxide based sensor has also been developed for continuous monitoring of hydrogen levels in argon cover gas. An electrochemical carbon sensor using a molten carbonate electrolyte and an oxygen sensor based on yttria doped thoria oxide electrolyte are also under advanced stage of development for measuring carbon and oxygen levels in sodium. Materials chemistry issues involved in developing these sensors and their operational experience in sodium system are highlighted in this presentation

  7. Method of advancing research and development of fast breeder reactors

    International Nuclear Information System (INIS)

    In the long term plan of atomic energy development and utilization, fast breeder reactors are to be developed as the main of the future nuclear power generation in Japan, and when their development is advanced, it has been decided to positively aim at building up the plutonium utilization system using FBRs superior to the uranium utilization system using LWRs. Also it has been decided that the development of FBRs requires to exert incessant efforts for a considerable long period under the proper cooperation system of government and people, and as for its concrete development, hereafter the deliberation is to be carried out in succession by the expert subcommittee on FBR development projects of the Atomic Energy Commission. The subcommittee was founded in May, 1986, to deliberate on the long term promotion measures for FBR development, the measures for promoting the research and development, the examination of the basic specification of a demonstration FBR, the measures for promoting international cooperation, and other important matters. As the results of investigation, the situation around the development of FBRs, the fundamentals at the time of promoting the research and development, the subjects of the research and development and so on are reported. (Kako, I.)

  8. Thermodynamics of ceramic breeder materials for fusion reactors

    International Nuclear Information System (INIS)

    Based on known or deduced phase relationships in ternary lithium oxygen systems such as Li-Al-O, Li-Si-O and Li-Zr-O, the unknown free enthalpy of formation values of ternary compounds are calculated starting from the known data of the compounds of the binary border systems. Criterion for the data assessment is interconsistency of the data of all the compounds within a given multi-component system. With the help of these data the development of partial pressures during the breeding process can be calculated for all the compounds of interest. In order to facilitate a compatibility assessment the quaternary systems Cr-Li-Si-O, Fe-Li-Si-O and Be-Li-Si-O were also investigated and thermodynamic data of pertinent ternary and quaternary compounds determined. Both the partial pressure development and the compatibility behaviour of a lithium containing compound are criteria for its qualification as a breeder material for a fusion reactor. (orig.)

  9. Multimegawatt Space Reactor Safety

    International Nuclear Information System (INIS)

    The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed

  10. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    International Nuclear Information System (INIS)

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

  11. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  12. Defect assessment procedure: A french approach for fast breeder reactors

    International Nuclear Information System (INIS)

    As a result of a collaborative effort between Commissariat a l'Energie Atomique, Electricite de France, and NOVATOME to produce and improve rules for fast breeder reactors, RCC-MR, an interim defect assessment procedure is now available in the first draft version (appendix A16). This procedure addresses defects detected during in-service inspection for reactor components operating at moderate or high temperature conditions. Three stages have been considered: initiation, propagation under cyclic loading with or without holdtime and crack instability by ductile and creep rupture. For each of these topics, procedures and rules based on fracture mechanics are proposed. Prediction of initiation is obtained by a simplified method named σd method which relies on the evaluation of the real stress-strain history on a small distance d (d = 0.05 mm for 316L(N) austenitic steel) close to the crack front and material characteristics (limiting stresses) that are available in nuclear codes. This method has been developed for fatigue, creep and creep-fatigue conditions. Defect growth assessment is performed for fatigue and creep-fatigue conditions. For creep-fatigue conditions, fatigue and creep crack growth per cycle are calculated separately and the total crack extension is taken as the sum of the two contributions. Extensive use of simplified method for estimating J (Js method) is made and developed when mechanical and thermal loadings are specified. On the final defect size, assessment may be made in order to avoid crack instability by ductile and creep rupture and collapse load on the remaining. The organization and contents of the present version of this appendix A16 is described. An overview of each specific rule is given

  13. SRP reactor safety evolution

    International Nuclear Information System (INIS)

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  14. Hydrodynamic and elastoplastic structural analysis of fast breeder reactor core accident

    International Nuclear Information System (INIS)

    This paper describes the principles and examples of applications of an explicit Lagrangian coupled finite difference-finite element code HEMP-ESI developed in order to calculate the structural consequences of hypothetical core disruptive accidents (HCDA) in nuclear reactors. The explicit solution algorithm of the finite difference scheme used to discretize the hydrodynamic fluid domains is shown to be very similar to that used for the solution of the finite element discretized shell structures, hence permitting an easy and efficient coupling. Two examples of simulation show the applicability of the method to nuclear reactor core safety analysis (test problem). Core explosion in a loop-type reactor including a shell containment: the calculation shows the energy absorbing function of the shell and enables the evaluation of the forces acting on the reactor containment. Hypothetical Core Disruptive Accident in a fast breeder reactor: the calculation shows the main features of this accident: lifting of the liquid sodium above the explosion and impact on the cover head inducing upward deformations; radial outflow of the sodium which induces large deformations of the inner and outer shell; zones of compressive circumferential stresses in the main shell at the junction of the spherical head and the cylindrical part

  15. Development of advanced in-service inspection technologies for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Periodic in-service inspection (ISI) is a mandatory requirement for all nuclear power facilities due to the steadily increasing obligations to provide evidence that plant and equipment is consistently satisfying safety and integrity requirements. Periodic ISI will ensure that service-induced defects and abnormalities are detected and remedied at an early stage. Further, the ISI data will be immensely useful for arriving at appropriate decision on the life extension of the plants. In view of the high radiation and temperature prevailing in the nuclear plants, customized remote handling technologies coupled with robotic devices are a must for carrying out ISI of critical components in these plants. The 500 MWe Prototype Fast Breeder Reactor(PFBR), which is under construction at Kalpakkam, Tamil Nadu, India is a pool type reactor comprising of main vessel, safety vessel, and reactor roof structure confining the primary coolant and cover gas with any associated radioactivity. PFBR is designed with provisions for continuous monitoring and surveillance of the structures forming the primary containment. Supplementing the continuous surveillance and monitoring, the integrity of the reactor vessels has to be assessed periodically by ISI during the reactor shutdown conditions and a comprehensive ISI system capable of functioning at 150 deg C has been conceived and formulated to meet this mandatory requirement. The comprehensive ISI system comprises of two remote-controlled four-wheeled robotic devices with non-destructive examination modules for volumetric and visual examinations and each device has a location-specific inspection requirements. The remote-controlled devices can move around the annulus of the main vessel and safety vessel with inspection modules for carrying out the inspection. An essential feature of the ISI of the PFBR vessels is the formulation and establishment of permanent reference markers on the MV and SV to identify the location of the ISI device in

  16. Current status of safety design and safety analysis for China ITER helium coolant ceramic breeder test blanket system long

    International Nuclear Information System (INIS)

    Helium Coolant Ceramic Breeder (HCCB) Test Blanket System (TBS) designed by China are planned to be tested in ITER to validate key technologies, including demonstration of nuclear safety, for future fusion reactor breeding blankets. Furthermore, in order to be operated in ITER, a nuclear facility (INB) recognized by French nuclear safety authority, safety design and safety analysis of the TBS are mandatory for the licensing procedures. This paper summarizes the status at current design phase with following main elements: The main radiological source terms in the system are tritium and activation products. Nuclear and tritium analysis are performed to identify their inventories and distributions in system. Multiple confinement barriers are considered to be the most essential safety feature. French regulation for pressure equipment and nuclear equipment (ESP/ESPN regulations) will be followed to ensure the system integrities. ALARA principle is kept in mind during the whole safety design phases. Protective actions including choice of advanced materials, improvement of shielding, optimization of operation and maintenance activities, usage of remote handling operations, zoning and access control have been considered. Passive safety is emphasized in the system design, only minimal active safety functions including call for fusion plasma shutdown and isolation of TBM from ex-vessel ancillary systems. High reliability and redundancies are required for components related to these functions. Several accidents have been identified and analyzed. Consider the limited inventories in the system and the intrinsic safety of fusion device, positive conclusions have been obtained. (author)

  17. Safety aspects of an inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Releases into the environment of radioactive materials contained in heavy ion fusion (HIF) reactor plants must be prevented by similar safety design concepts as they are applied to present fission converter (e.g. LWR's) and breeder reactors (LMFBR's). This study is intended to identify significant safety aspects of inertial confinement fusion power plant concepts and to relate them to the more familliar basis of knowledge about the safety and the hazards of other advanced nuclear power reactor systems such as the LMFBR. Needs for safety related research and development specifically for inertial confinement fusion are pointed out. (orig./GG)

  18. Safety of research reactors

    International Nuclear Information System (INIS)

    The number of research reactors that have been constructed worldwide for civilian applications is about 651. Of the reactors constructed, 284 are currently in operation, 258 are shut down and 109 have been decommissioned. More than half of all operating research reactors worldwide are over thirty years old. During this long period of time national priorities have changed. Facility ageing, if not properly managed, has a natural degrading effect. Many research reactors face concerns with the obsolescence of equipment, lack of experimental programmes, lack of funding for operation and maintenance and loss of expertise through ageing and retirement of the staff. Other reactors of the same vintage maintain effective ageing management programmes, conduct active research programmes, develop and retain high calibre personnel and make important contributions to society. Many countries that operate research reactors neither operate nor plan to operate power reactors. In most of these countries there is a tendency not to create a formal regulatory body. A safety committee, not always independent of the operating organization, may be responsible for regulatory oversight. Even in countries with nuclear power plants, a regulatory regime differing from the one used for the power plants may exist. Concern is therefore focused on one tail of a continuous spectrum of operational performance. The IAEA has been sending missions to review the safety of research reactors in Member States since 1972. Some of the reviews have been conducted pursuant to the IAEA' functions and responsibilities regarding research reactors that are operated within the framework of Project and Supply Agreements between Member States and the IAEA. Other reviews have been conducted upon request. All these reviews are conducted following procedures for Integrated Safety Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety

  19. The present status of the fast breeder reactor industrialization in western Europe

    International Nuclear Information System (INIS)

    The development of the liquid metal fast breeder reactor in Europe started in the mid-fifties, after the successful operation of EBR-1 at ARCO, Idaho, in 1951. A more and more integrated development among the countries of the European Community culminated in 1986 with the beginning to power of the 1200 MWe SUPERPHENIX plant at Creys-Malville, France. The road is now open towards the full industrialization of the liquid metal fast breeder reactor at the moment, in 2005, when the first European thermal neutron power reactor station will have to be decommissioned and replaced. The European programme aims at providing the utilities at that time with a clear choice between thermal neutron reactors and fast breeder reactors, both economical but very different in their use of the limited natural resource that uranium is. (author)

  20. Present status of the fast breeder reactor industrialization in western Europe

    International Nuclear Information System (INIS)

    The development of the liquid metal fast breeder reactor in Europe started in the mid-fifties, after the successful operation of EBR-1 at ARCO, Idaho, in 1951. A more and more integrated development among the countries of the European Community culminated in 1986 with the startup of the 1200 MWe SUPERPHENIX plant at Creys-Malville, France. The road is now open towards the full industrialization of the liquid metal fast breeder reactor at the moment, in 2005, when the first European thermal neutron power reactor station will have to be decommissioned and replaced. The European programme aims at providing the utilities at that time with a clear choice between thermal neutron reactors and fast breeder reactors, both economical but very different in their use of the limited natural resources that uranium

  1. Evaluation of symbiotic energy system between gas-cooled fast breeder reactor (GCFR) and multi-purpose very high temperature reactor (VHTR), (4)

    International Nuclear Information System (INIS)

    The conceptual design study of 1000 MWe gas-cooled fast breeder reactor (GCFR), which is used in the GCFR-VHTR symbiotic energy system, has been performed. In this report, the transient response of the GCFR core to accident events has been analyzed and safety performance of the 1000 MWe GCFR has been evaluated considering the analyses. A depressurization accident caused by failure of a primary coolant system and a reactivity insertion accident due to withdrawal of a control rod have been analyzed using nuclear and thermo-hydraulic coupled program MR-X developed for kinetics analysis of gas-cooled fast breeder reactors. The maximum fuel and cladding temperatures are most important problem to be analysed during a trangient of a gas-cooled fast breeder reactors. The analyses show that reliable reactor shutdown and emergency cooling systems are most important to achieve successful cold shutdown well before leading to core damage and also that no severe failures of fuel pin and cladding occures by working above mentioned safety systems well during the accidents. (author)

  2. Fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs)

    International Nuclear Information System (INIS)

    It is intended that this review will be useful not only to scientists but also to those concerned with design, day-to-day operation of plant, with liquid metal fast breeder reactors (LMFBRs), safety and decommissioning. Because of this, the review has been widened to include not only the mass transfer behaviour of the various radionuclides in experimental and operating systems, but also the monitoring of the various species, the methods of measurement and the development of methods to control the build-up of the more important long half-life species in operating plants. The information used in the review has been taken from open literature sources to provide an up-to-date presentation of the behaviour of the various isotopes in LMFBRs. 172 refs, 14 figs, 22 tabs

  3. Numerical simulation of sodium pool fires in liquid metal-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    In Liquid Metal-Cooled Fast Breeder Reactor (LMFBR), the leakage of sodium can result in sodium fires. Due to sodium's high chemical reactivity in contact with air and water, sodium fires will lead to an immediate increase of the air temperature and pressure in the containment. This will harm the integrity of the containment. In order to estimate and foresee the sequence of this accident, or to prevent the accident and alleviate the influence of the accident, it is necessary to develop programs to analyze such sodium fire accidents. Based on the work of predecessors, flame sheet model is produced and used to analyze sodium pool fire accidents. Combustion model and heat transfer model are included and expatiated. And the comparison between the analytical and experimental results shows the program is creditable and reasonable. This program is more realistic to simulate the sodium pool fire accidents and can be used for nuclear safety judgement. (authors)

  4. Gas-cooled fast breeder reactor shielding benchmark calculation

    Energy Technology Data Exchange (ETDEWEB)

    Rouse, C.A.; Mathews, D.R.; Koch, P.K.

    1977-01-01

    This report summarizes the results of a shielding benchmark calculation performed by General Atomic (GA) and Oak Ridge National Laboratory (ORNL). The problem analyzed was a neutron-coupled gamma ray transport calculation of the core blanket shield of the 300-MW(e) gas-cooled fast breeder reactor (GCFR). Comparison of the initial GA and ORNL results indicated good agreement for fast fluxes (E greater than 0.9 MeV and E greater than 0.086 MeV) but poor agreement for epithermal and thermal neutron fluxes. Examination of the results revealed that a deficiency in the GA fine-group cross section preparation code was responsible for the differences in the GA and ORNL iron cross sections. Modification of the GA cross sections to include self-shielding was accomplished, and the updated GA benchmark calculation performed with the self-shielded iron cross sections was in excellent agreement with the ORNL results for fast neutron fluxes with E greater than 0.9 MeV and E greater than 0.086 MeV and in good agreement for epithermal and thermal fluxes. The agreement of the gamma heating rates also improved significantly. Thus, it was concluded that the good agreement of the GA and ORNL neutron-coupled gamma ray transport calculation indicates that (1) the methods and cross sections used by both laboratories were compatible and consistent and (2) the use of 24 neutron energy groups and 15 gamma energy groups by GA was adequate compared with the use of 51 neutron energy groups and 25 gamma energy groups by ORNL.

  5. Network Representation of Design Knowledge of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    A method of design knowledge representation was studied for the Japanese fast breeder reactor Monju, aiming at enhanced understanding of engineering considerations with mutual relations. Taking over design knowledge of Monju to next generation designers/engineers to be in charge of design of future FRs is by no means easy, in contrast with operation and maintenance knowledge which can be acquired in the real plant operation and maintenance. Specifications of the as-is Monju contains only a small part of the entire design knowledge, mainly by two reasons. Firstly, reasons for selecting the as-is specifications can not be understood until reaching proper knowledge source. Secondly, there are many rejected options on the design specifications. Design specifications are selected along with technical dependencies among a huge number and diversified specification items. Decisions design are made basically along with these dependencies which can hardly be traced in the currently available database or document libraries. Reasons for the rejections of options need to be profoundly understood, because those are not certainly due to technical inferiority. Some of rejected options can be worth reconsidering in the future, possibly by technical advances in materials, high-precision prediction software tools, rationalized standards/code, etc. The authors propose a new design knowledge representation approach based on networking of knowledge nodes along with the mutual dependencies. A prototype software has been developed and a basic performance test was made to visualize the dependency network. An additional function to enable design case studies on hypothetical adoptions of rejected options is now under consideration. (author)

  6. Argon entrainment into liquid sodium in fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: ► In the present work, different types of entrainment mechanisms have been studied. ► Onset of entrainment could be characterized with turbulent intensities. ► For vortex type entrainment, a correlation has been proposed. - Abstract: Gas entrainment in pool type sodium cooled fast breeder reactors has been a subject of great interest for a quite long time now. The issue of entrainment of argon cover gas in LMFBR's is being addressed by fundamental studies. Present work focuses on characterization of onset of shear type entrainment and liquid fall type entrainment based on mean velocity and turbulent kinetic energy at liquid surface. Study also includes characterization of onset of vortex type entrainment based on mean velocities (time averaged) in the outlet pipe. Experiments were carried out to characterize shear type entrainment in stirred tank with different impeller geometries with air–water and xylene–water systems. Onset of liquid fall type entrainment was studied with cylindrical tank with a nozzle whose input angle varied. Mean and r.m.s. velocity profiles near the liquid surface were measured with the help of ultrasonic velocity profiler (UVP). The results are compared with other literature. It is observed that the onset of entrainment can be characterized by the turbulent kinetic energy near the free liquid surface. Re-submergence angle was measured and r.m.s. velocities found to be in the same range as in case of shear type of entrainment. Cylindrical tank with tangential inlet and bottom outlet was used to study onset of vortex formation. Effect of different parameters like outlet diameter, tank diameter and liquid height in the tank on critical velocity was studied and correlation has been proposed.

  7. Fast breeder reactors in relation to energy requirements. Chapter 2.2

    International Nuclear Information System (INIS)

    It is shown that the world is going to need substantial quantities of energy from new sources early in the next century. Although it may be possible to get a significant amount from solar and geothermal sources, it is far too early to predict how large a contribution they will be able to make. Nuclear power can make a large contribution and it would be wrong to do anything to close this option at this stage. Although there is considerable uncertainty about the precise quantity of commercially exploitable uranium in the world, it is almost certain that breeder reactors will be required. The time-scale is such that utilities throughout the world will need to be able to order breeders in quantity with complete confidence for operation from the mid 1990s onwards. The engineering, safety and logistic problems that are considered in the other chapters of this book are such that, if we are to meet this time-scale, we must press ahead with the utmost urgency. (author)

  8. Development of high nitrogen electrodes for fast breeder reactor applications

    International Nuclear Information System (INIS)

    Austenitic stainless steels of AISI type 316 (316 SS) and its variants are used extensively as structural material for the components of fast reactors operating at temperature up to 823 K. SS 316LN has been chosen as the major structural material for the construction of Prototype Fast Breeder Reactor (PFBR) with a targeted service life of 40 years. To reduce the risk of sensitization in SS 316LN, the carbon content has been reduced to less than 0.03 wt%, and the nitrogen content has been specified as 0.08 wt% to compensate the loss in strength due to the reduced carbon content. An improved version of this alloy with nitrogen content of 0.14 wt% in a frilly austenite matrix has been developed for the future FBRs, to enhance the service life of the structural components up to 60 years. Indigenously developed modified E3 16-1 5 electrodes were used for the fabrication of PFBR components to enhance the structural reliability of the components. The modifications from AWS/ASME SFA 5.4 include stringent composition limits, narrow range of ferrite content, and impact toughness after aging at 1023K for 100h, tensile properties at elevated (service) temperatures and intergranular corrosion (IGC) test as per ASTM A262 Practice E. Since the improved version alloy is rich in nitrogen content than the existing alloy, it has become necessary to develop a welding consumable with reasonably good weldability that is suitable for the fabrication of future FBR components. At present there are no commercially available welding consumables to weld these steels and the development is under way. In this work, a matching consumable methodology was adopted to develop the welding consumable. However, as per specification targeting the chemistry, solidification mode and delta ferrite was challenging, since the solidification mode of the weld metal shifts to fully austenitic region due to dilution of nitrogen from the base metal, which may increase the risk of hot cracking susceptibility

  9. Status of fast breeder reactor development in India

    International Nuclear Information System (INIS)

    The energy scenario and economic conditions in India are presented. India needs considerable energy for its rapid industrialisation with the liberal economic policy. Nuclear energy with FBR is the only large scale energy resource other than coal, available in the country. The present economic constraints have delayed the construction of new NPPs. The performance of operating reactors has improved considerably during the year. Operating experience of FBTR has been detailed particularly the reactivity incident and its investigations. Updated design of 500 MWe PFBR is presented. Various R and D works in support of FBR in the engineering, metallurgy, chemistry, reprocessing, safety etc. are detailed. (author)

  10. Seismic analysis of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    This report is a general survey of the recent methods to predict the seismic structural behaviour of LMFBRs. It shall put into evidence the impact of seismic analysis on the design of the different structures of the reactor. This report is addressed to specialists and institutions of governmental organizations in industrialized and developing countries responsible for the design and operation of LMFBRs. The information presented should enable specialists in the R and D institutions and industries likely to be involved, to establish the correct course of the design and operation of LMFBRs. Also, the safety aspect of seismic risk are emphasized in the report. Refs and figs

  11. Fast breeder reactors insertion in a D2O - natural U nuclear power plants park

    International Nuclear Information System (INIS)

    A model for the evolution of Argentine's installed nuclear power for the next 40 years is presented. The consequences of fast breeder reactors' introduction are studied in both autarchic Pu cycle and a limited reprocessing system. The passage of a reactor park like the national, of natural U - heavy water to one of fast breeder reactors, can only be obtained in a very long term due, fundamentally, to the need of Pu produced for those to feed the last ones. (M.E.L.)

  12. Current status of development of Demonstration Fast Breeder Reactor and prospect of FBR commercialization

    International Nuclear Information System (INIS)

    The Demonstration Fast Breeder Reactor (DFBR) is the next step of FBR development following the prototype fast breeder reactor 'MONJU'. The DFBR is now under development by The Japan Atomic Power Co. (JAPC) under the sponsorship of 9 Japanese electric power companies and Electric Power Development Co., Ltd. The JAPC has been performing the design study and R and D for DFBR in cooperation with Power Reactor and Nuclear Fuel Development Corp. (PNC), Central Research Institute of Electric Power Industry (CRIEPI) and Japan Atomic Energy Research Institute (JAERI). This report describes the prospect of FBR commercialization and the current status of new technology for DFBR and innovative technology FBR commercialization. (author)

  13. Production of nuclear fusion reactor fuel by ceramic tritium breeder material

    International Nuclear Information System (INIS)

    Fuel tritium is generated from the nuclear reaction between the fusion neutron and the lithium of the breeder material arranged in the blanket that encloses the fusion plasma in the fusion reactor. However, the release process of the generated tritium has not been completely clarified. Recently, Japan Atomic Energy Agency started the tritium generation and recovery experiment in using nuclear fusion neutron source (FNS). In this report, the recent results of study on breeder material and its manufacturing technology is presented. (author)

  14. Y2K issues for real time computer systems for fast breeder test reactor

    International Nuclear Information System (INIS)

    Presentation shows the classification of real time systems related to operation, control and monitoring of the fast breeder test reactor. Software life cycle includes software requirement specification, software design description, coding, commissioning, operation and management. A software scheme in supervisory computer of fast breeder test rector is described with the twenty years of experience in design, development, installation, commissioning, operation and maintenance of computer based supervision control system for nuclear installation with a particular emphasis on solving the Y2K problem

  15. Fusion reactor safety

    International Nuclear Information System (INIS)

    Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

  16. Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

  17. Welding and reactor safety

    International Nuclear Information System (INIS)

    The high safety requirements which must be demanded of the quality of the welded joints in reactor technique have so far not been fulfilled in all cases. The errors occuring have caused considerable loss of availability and high material costs. They were not, however, so serious that one need have feared any immediate danger to the personnel or to the environment. The safety devices of reactor plants were only called upon in a few cases and to these they responded perfectly. The intensive efforts to complete and improve the specifications are to contribute to that in future, the reactor plants can be counted even more so as one of the safest technical plants ever. (orig./LH)

  18. 03 - Sodium cooled fast breeder fourth-generation reactors - The technological demonstrator ASTRID

    International Nuclear Information System (INIS)

    After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

  19. Characteristics of fission products behavior on a severe accident in fast breeder reactor

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) has been developing the ACTOR code for the analysis of the fission products behavior under the severe accident condition to apply the probabilistic safety assessment to fast breeder reactor plants. Major analysis models of the ACTOR code were validated and adjusted by related experimental results. The fission products behavior on PLOHS (Protected Loss of Heat Sink) sequence which is one of the typical severe accidents in FBR plant was analyzed by using the ACTOR code. It was confirmed that the ACTOR had an enough capability to analyze the fission products behavior during severe accident. From the analysis results of PLOHS, it was confirmed that cesium is transferred to the cover gas region much more than iodine because iodine which is one of halogen connects to sodium easily and is retained in the coolant. Therefore, cesium is important and it is needed to examine the necessity to treat cesium as one of FPs considered in reactor establishment permission for FBR plant. Thus, cesium transfer behavior in sodium during the rare gas bubbles rise from fuel to the cover gas region was confirmed to be very important. And JNES started study including validation test about cesium transfer behavior with Hokkaido University. (author)

  20. Modeling and Simulation of Operator Training Simulator for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Often the challenge faced by the Nuclear Power Industry is the availability of well trained human resource for efficient power plant monitoring and control. Safety of the plant purely depends upon the plant knowledge acquired, understanding of plant dynamics and the skills possessed by the operators through systematic training. Generally the operators are given class room and field training before deploying them in the operation crew. But, for handling emergency and abnormal conditions, the formal class room training and field training have proved to be inadequate according to the expert estimates. The state- of- art computer based operator training simulators covering the full spectrum of the plant have become an essential element in bridging the gap between the inadequacy and efficiency. Gradually the training simulators are getting embedded in the operator training programme and started playing a crucial role in enhancing the ability of the operators.This paper discusses about the operator training simulator called KALBR- SIM i.e. Kalpakkam Breeder Simulator that has been built at IGCAR for training the Prototype Fast Breeder Reactor (PFBR) operators. It is a Full Scope Replica Operator Training Simulator built to replicate PFBR. The scope of the paper covers the basic modules necessary for building each process model of the simulator, design and development of the reactor sub systems like Neutronics, Primary Sodium, Secondary Sodium, Decay Heat Removal, Steam Water, Electrical systems and the associated logics and controls. It is followed by a detailed discussion on replication aspects of Simulator Control Room and its advantages, the Hardware Architecture, Instruction Station facility and loading of scenarios. It further elaborates on Steady State and Bench Mark Transients tests conducted on the Operator Training Simulator like One primary sodium pump trip, one primary pump seizure, Primary pipe rupture, one boiler feed pump trip and Station Black Out. (author)

  1. IAEA note on multi-national fuel cycle centres as related to fast breeder reactors

    International Nuclear Information System (INIS)

    The significant aspects of associating fast breeder reactor fuel cycles with the concept of regional fuel cycle centres, as studied earlier by the IAEA, are identified. The results of the RFCC Study Project are presented, and how in particular non-proliferation and safeguards, radioactive waste management and economic considerations would be effected by inclusion of fast breeder reactor fuel cycle facilities and possibly fast breeder reactors as well in such centres, are discussed. The current effort of the IAEA to develop a computer programme which models the material flows in the nuclear fuel cycle which could be applied to the analysis of alternative siting strategies for FBR and its fuel cycle facilities is discussed

  2. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  3. Perspectives on reactor safety

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  4. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  5. The Future of Nuclear Energy: Facts and Fiction Chapter IV: Energy from Breeder Reactors and from Fusion?

    OpenAIRE

    Dittmar, Michael

    2009-01-01

    The accumulated knowledge and the prospects for commercial energy production from fission breeder and fusion reactors are analyzed in this report. The publicly available data from past experimental breeder reactors indicate that a large number of unsolved technological problems exist and that the amount of "created" fissile material, either from the U238 --> Pu239 or from the Th232 --> U233 cycle, is still far below the breeder requirements and optimistic theoretical expectations. Thus huge e...

  6. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  7. Reactor safety systems

    International Nuclear Information System (INIS)

    The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.)

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

    International Nuclear Information System (INIS)

    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

  9. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  10. MINT research reactor safety program

    Energy Technology Data Exchange (ETDEWEB)

    Mohamad Idris bin Taib [Division of Special Project, Malaysian Institute for Nuclear Technology Research (MINT), Bangi (Malaysia)

    2000-11-01

    Malaysian Institute for Nuclear Technology Research (MINT) Research Reactor Safety Program has been done along with Reactor Power Upgrading Project, Reactor Safety Upgrading Project and Development of Expert System for On-Line Nuclear Process Control Project. From 1993 up to date, Neutronic and Thermal-hydraulics analysis, Probabilistic Safety Assessment as well as installation of New 2 MW Secondary Cooling System were done. Installations of New Reactor Building Ventilation System, Reactor Monitoring System, Updating of Safety Analysis Report and Upgrading Primary Cooling System are in progress. For future activities, Reactor Modeling will be included to add present activities. (author)

  11. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  12. Status of national programmes on fast breeder reactors. Eighteenth annual meeting, Vienna, Austria, 16-19 April 1985

    International Nuclear Information System (INIS)

    The Eighteenth Annual Meeting on the Status of National Programmes in Member States of the IAEA on Fast Breeder Reactors had been held in April 1985. The representatives of the Member States and international organizations reported status and activities in the field of fast breeder reactors development and operation. A separate abstract was prepared for each of the 12 presentations of the meeting

  13. Analysis of passive shutdown capability for a loss of flow accident in a medium sized liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    The passive shutdown capability of a medium sized (500 MWe) liquid-metal fast breeder reactor with oxide, carbide and metal fuels has been analysed for a loss of flow accident using static and dynamic analysis methods. The carbide fuel is assumed to be He-bonded as well as Na-bonded. The dependence of the passive safety on the flow halving time constant of the loss of flow incident and the feedback components, like radial core expansion due to subassembly spacer pad heating and differential control rod expansion due to heating of the control rod suspension mechanism, is highlighted. (author)

  14. Remote fabrication of pellet fuels for United States breeder reactors

    International Nuclear Information System (INIS)

    Goal of the program is to demonstrate the feasibility of fabricating breeder fuel in a remotely operated and maintained mode by 1985. Development for pellet fuel fabrication is in the engineering stage with much of the equipment for ceramic unit operations in final design or currently under testing. Results to date confirm that remote fabrication of pellet fuels is feasible. Several of the processes and equipment items are described in this report

  15. NDT services of Prototype Fast Breeder Reactor PFBR - a contribution from Blue Star Ltd towards growth

    International Nuclear Information System (INIS)

    Prototype fast breeder reactor is 500MWe sodium cooled reactor of pool type design. Three classes of steels namely austenitic stainless steel (304LN and 316LN), Ferritic steel (Modified 9Cr 1 Mo) and Carbon steel materials (A48P2) were used for manufacturing of different nuclear steam supply system (NSSS) components. The components are in different product forms such as castings, forgings, plates, rounds, Hollow bars, Seamless tubes and pipes for fitness-for- purpose applications. Due to the criticality of the design, stringent quality control measures were to be adopted for component integrity. The collaboration of Blue Star Ltd with major players of fabrication such as M/s. Larsen and Toubro Ltd (main and safety vessels, sodium tanks), Kirloskar Brothers (primary sodium pump forgings), Walchand Nagar Industries (grid plate assembly), Bharat Heavy Electricals (inner vessel and thermal baffle), MTAR Technologies (grid plate sleeves) and Omplas Systems (colmonoy hard facing of various core components) had witnessed many challenges in achieving the required quality. Different code based (RCCMR and ASME) approaches for acceptance criteria had to be adopted for evaluation of the components that are large in dimensions and quantities. This paper discusses the knowledge gained through different procedural developments for these critical components and provided suggestion of remedial measures in achieving the required quality. (author)

  16. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    Science.gov (United States)

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation. PMID:25725884

  17. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    Science.gov (United States)

    Malathi, N.; Sahoo, P.; Ananthanarayanan, R.; Murali, N.

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  18. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malathi, N.; Sahoo, P., E-mail: sahoop@igcar.gov.in; Ananthanarayanan, R.; Murali, N. [Real Time Systems Division, Electronics, Instrumentation and Radiological Safety Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2015-02-15

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

  19. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    International Nuclear Information System (INIS)

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation

  20. Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Highlights: • This work analyzes important trends of the running-in phase of a thorium breeder PBR. • Depletion equations are solved for important actinides and a fission product pair. • Breeding U-233 is achieved in 7 years by cleverly adjusting the feed fuel enrichment. • A safety analysis shows the thorium PBR is passively safe during the running-in phase. - Abstract: The present work investigates the running-in phase of a 100 MWth Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated. For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme. Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300 days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7 years after start of reactor operation. The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the

  1. Compilation of data and descriptions for United States and foreign liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    This document is a compilation of design and engineering information pertaining to liquid metal cooled fast breeder reactors which have operated, are operating, or are currently under construction, in the United States and abroad. All data has been taken from publicly available documents, journals, and books

  2. Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

    International Nuclear Information System (INIS)

    Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections

  3. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

  4. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  5. Atoms in Appalachia. Historical report on the Clinch River Breeder Reactor site

    Energy Technology Data Exchange (ETDEWEB)

    Schaffer, D

    1982-01-01

    The background information concerning the acquisition of the land for siting the Clinch River Breeder Reactor is presented. Historical information is also presented concerning the land acquisition for the Oak Ridge facilities known as the Manhattan Project during World War II.

  6. Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.W.; Baes, C.F. III; Dunning, D.E. Jr.

    1980-05-01

    Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections.

  7. ORIGEN2 model and results for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A G; Bjerke, M A

    1982-06-01

    Reactor physics calculations and literature information acquisition have led to the development of a Clinch River Breeder Reactor (CRBR) model for the ORIGEN2 computer code. The model is based on cross sections taken directly from physics codes. Details are presented concerning the physical description of the fuel assemblies, the fuel management scheme, irradiation parameters, and initial material compositions. The ORIGEN2 model for the CRBR has been implemented, resulting in the production of graphical and tabular characteristics (radioactivity, thermal power, and toxicity) of CRBR spent fuel, high-level waste, and fuel-assembly structural material waste as a function of decay time. Characteristics for pressurized water reactors (PWRs), commercial liquid-metal fast breeder reactors (LMFBRs), and the Fast Flux Test Facility (FFTF) have also been included in this report for comparison with the CRBR data.

  8. Report of the Institute for Reactor Safety on research and development activities in 1994

    International Nuclear Information System (INIS)

    Contributions to nuclear safety research concerned the design basis and safety of innovative reactor concepts as well as unsolved problems in the late phase of a core melt-through accident at future light-water-moderated reactors. The works regarding the breeder reactor were finalized. Contributions to the project of nuclear fusion focussed on the development of breeding blankets, high-resistance components, vacuum engineering and proof of the safety of components and systems as well as blanket engineering. (HP)

  9. Fuel pins and core response under liquid-metal fast breeder reactor transient overpower accident conditions

    International Nuclear Information System (INIS)

    Since the earlier liquid-metal fast breeder reactor transient overpower assessments were done (1975), new experimental data and modeling improvements have occurred that indicate later failures and more molten fuel squirted into the channel with a higher propensity for plugging. An initial sweepout still occurs, and an analysis shows that even if coherent instead of the expected stochastic failures occur, the blockages are partial, the reactor is strongly shut down, and a coolable geometry exists. Hence, the overall consequences would be benign

  10. ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

    International Nuclear Information System (INIS)

    Highlights: • Metal fuel FBR safety can be assessed by its response to unprotected transients. • Safety during unprotected loss of flow accident (ULOF) is important for FBR cores. • ULOF analyses are carried out as a function of core size and perturbation method. • Smaller metal cores are found to be safer with respect to the ULOF accidents. • 1st order perturbation method gives conservative results in an ULOF accident. - Abstract: The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120 MWe, 500 MWe and 1000 MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase

  11. ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

    Energy Technology Data Exchange (ETDEWEB)

    Riyas, A., E-mail: rias@igcar.gov.in [111B, CDO, Reactor Physics Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Mohanakrishnan, P. [Adjunct Professor, Manipal University, Manipal (India)

    2014-10-15

    Highlights: • Metal fuel FBR safety can be assessed by its response to unprotected transients. • Safety during unprotected loss of flow accident (ULOF) is important for FBR cores. • ULOF analyses are carried out as a function of core size and perturbation method. • Smaller metal cores are found to be safer with respect to the ULOF accidents. • 1st order perturbation method gives conservative results in an ULOF accident. - Abstract: The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120 MWe, 500 MWe and 1000 MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase.

  12. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored

  13. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  14. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  15. Advances in reactor safety research

    International Nuclear Information System (INIS)

    The Nuclear Safety Project is an important part of the German reactor safety research programme. It works on problems concerning safety and environemental risks of LWR reactors and reprocessing plants and investigates accident consequences. At the 1978 annual meeting, the core behaviour on cooling and reactivity disturbances was discussed, as well as release, retention, and possible radiological effects of radioactive pollutants. Among other subjects, fission product retention in LWR reactors and reprocessing plants were reported on as well as hypothetic core meltdown. (orig.)

  16. Status of National Programmes on Fast Breeder Reactors. International Working Group on Fast Reactors Twenty-First Annual Meeting, Seattle, USA, 9-12 May 1988

    International Nuclear Information System (INIS)

    The following papers on the status of national programmes on fast breeder reactors are presented in this report: Fast breeder reactor development in France during 1987; Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands; A review of the Indian fast reactor programme; A review of the Italian fast reactor programme; A review of the fast reactor programme in Japan; Status of fast reactor activities in the USSR; A review of the United Kingdom fast reactor programme; Status of liquid metal reactor development in the United States of America; Review of activities of the Commission of European Communities relating to fast reactors in 1987; European co-operation in the field of fast reactor research and development — 1987 progress report; A review of fast reactor activities in Switzerland

  17. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. A review. Pt. I. Key areas

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions, Coimbatore, Tamilnadu (India)

    2013-11-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or ready-made supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution. (orig.)

  18. Reactor safety - an international task

    International Nuclear Information System (INIS)

    The dimensions and the significance of the task of ensuring reactor safety can be defined on the basis of experiences gained from Harrisburg and Chernobyl. The countries that use nuclear energy are tied together to a community by virtue of the risk they share. Therefore the GRS is working in close cooperation with the EC, OECD, IAEO and COMECON. This results in safety examinations of the Greifswald reactor, safety analyses of nuclear reactors in Germany, France and the USA and also considerations on the safety demands to be placed on new reactor concepts. (DG)

  19. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    Science.gov (United States)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  20. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like – Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) and Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  1. Overview of pool hydraulic design of Indian prototype fast breeder reactor

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  2. An investigation of nuclear physics characteristics of fast breeder reactors (LMFBR and GCFBR) with various fuel cycles

    International Nuclear Information System (INIS)

    The primary emphasis on the study has been placed on comparing neutronic characteristics, e.g. fissile inventory, breeding and safety, of fast breeder reactors with uranium-plutonium and thorium-uranium fuel cycles. The study was performed using identical calculation methods and consistent data basis. As the reference fast breeder reactor, two different types of 1,200 MWe PuO2-UO2 fuelled fast reactors were chosen, which are sodium-cooled fast breeder reactor (LMFBR) and helium-cooled fast breeder reactor (GCFBR). The following four fuel utilisation models were investigated for each of LMFBR and GCFBR. (1) PuO2-UO2 core, and UO2 axial and radial blankets, (2) PuO2-UO2 core, UO2 axial blanket and ThO2 radial blanket, (3) 233UO2-UO2 core, and ThO2 axial and radial blankets, (4) 233UO2-ThO2 core, and ThO2 axial and radial blankets. The main results obtained are summarised as follows: (1) Pu fuelled LMFBR provides sufficiently high breeding gain, but has unfavourable characteristics of considerably large positive sodium-void reactivity effect. (2) U-233 fuelled LMFBR provides the favourable characteristics of negative sodium-void reactivity effect, but provides either negative or very low breeding gain. (3) Pu fuelled GCFBR has the desirable characteristics from the viewpoints investigated in the study, i.e. relatively low fissile inventory, very large breeding gain, sufficiently negative Doppler reactivity effect and negative steam ingress reactivity effect. (4) Use of U-233 in the core of GCFBR is not preferable, because of substantially low breeding gain and terribly large positive steam ingress reactivity effect. (5) Use of ThO2 in the core of LMFBR and GCFBR instead of UO2 leads to increase of fissile inventory and decrease of breeding gain. (6) Use of ThO2 in the blanket of LMFBR and GCFBR instead of UO2 does not give any significant influence on the neutronic characteristics

  3. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  4. Stress Analysis of Steam Generator Shell Nozzle Junction for Sodium cooled Fast Breeder Reactor

    Directory of Open Access Journals (Sweden)

    Mani N,

    2010-07-01

    Full Text Available The Steam Generators (SG decides the capacity factor in Sodium cooled Fast breeder Reactor (SFR plants and hence they are designed with high reliability. One of the critical locations in SG is the shell nozzle junction. This junction is subjected to an end bending moment and internal pressure. Since the shell nozzle junction is the critical location of SG a double-ended guillotine rupture will result in leakage of large quantity of sodium, which is not desirable. The material of construction is modified 9Cr-1Mo. Hence safety equirements demand that Leak Before Break criteria with assumed initial flaw have to be demonstrated. To demonstrate LBB, the basic requirement is to predict the state of stress precisely in the shell nozzle junction under various loading conditions. An efficient finiteelement modeling for shell nozzle junction has been presented in which shell elements are employed to idealize the whole region. These results are used for the analysis of leak before break concept.

  5. Power supply for control and instrumentation in Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    The design and operation of the four 'no-break' power supplies for control and instrumentation in the Fast Breeder Test Reactor (FBTR), Kalpakkam, are described. Interruptions in the power supplies are eliminated by redundancy and battery back-up source while voltage dips and transients are taken care by automatic regulation system. The four power supplies are : (1) 24 V D.C. exclusively for neutronic and safety circuits, (2) 48 V D.C. for control logic indication lamps and solenoid valves, (3) 220 V D.C. for switchgear control, control room emergency lighting and D.C. flushing oil pump for the turbine and (4) 220 V A.C. single-phase 50 H/Z for computers and electronics of control and instrumentation. Stationary lead-acid batteries (lead antimony type) in floating mode operation with rectifier/charger are used for emergency back-up. All these power supplies are fed by 415 V, 3-phase, 50 HZ emergency supply buses which are provided with diesel generator back-up. Static energy conversion system (in preference to mechanical rotation system) is used for A.C. to D.C. and also for A.C. to A.C. conversion. (M.G.B.)

  6. Phase 1 of feasibility studies on commercialized fast breeder reactor cycle system

    Energy Technology Data Exchange (ETDEWEB)

    Konomura, Mamoru [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2001-09-01

    The situation of the feasibility study on commercialized fast breeder reactor cycle system is reported. This is the joint study of JNC and major domestic electric power companies. The study intends to construct and propose the concept of candidate FBR cycle responding various needs in 21 century. The target of the study is to extract plural candidates of commercialized FBR cycle system including FBR plant, reprocessing facility and fuel fabrication facility and to propose realization scenario such as R and D schedule. The study settles on the comprehensive evaluation guideline on not only safety but also on economy, effective utilization of resources, reduction of environmental load and non-proliferation of nuclear materials. Based on the guideline, the design target items are established. During the Phase 1 (FY1999 and 2000), the extensive FBR systems are surveyed in the study. As coolant, Na heavy metals such as Pb, PbBi, gases such as CO2, He and water (H2O and D2O) are surveyed. As fuel, pin type such as oxide, metal and nitride, coated particle type of oxide and nitride and liquid fuel are considered combined with coolant. The final report of Phase 1 will be published shortly. In Phase 2 of the study starting FY2001 until FY2005, the extraction of the candidates will be performed. (K. Tsuchihashi)

  7. Phase 1 of feasibility studies on commercialized fast breeder reactor cycle system

    International Nuclear Information System (INIS)

    The situation of the feasibility study on commercialized fast breeder reactor cycle system is reported. This is the joint study of JNC and major domestic electric power companies. The study intends to construct and propose the concept of candidate FBR cycle responding various needs in 21 century. The target of the study is to extract plural candidates of commercialized FBR cycle system including FBR plant, reprocessing facility and fuel fabrication facility and to propose realization scenario such as R and D schedule. The study settles on the comprehensive evaluation guideline on not only safety but also on economy, effective utilization of resources, reduction of environmental load and non-proliferation of nuclear materials. Based on the guideline, the design target items are established. During the Phase 1 (FY1999 and 2000), the extensive FBR systems are surveyed in the study. As coolant, Na heavy metals such as Pb, PbBi, gases such as CO2, He and water (H2O and D2O) are surveyed. As fuel, pin type such as oxide, metal and nitride, coated particle type of oxide and nitride and liquid fuel are considered combined with coolant. The final report of Phase 1 will be published shortly. In Phase 2 of the study starting FY2001 until FY2005, the extraction of the candidates will be performed. (K. Tsuchihashi)

  8. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    International Nuclear Information System (INIS)

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators

  9. Control rod calibration methods for fast breeder reactors applied to Phenix

    International Nuclear Information System (INIS)

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

  10. Preliminary physics design of accelerator-driven thorium cycle fast breeder reactor

    International Nuclear Information System (INIS)

    A preliminary reactor physics design of a lead cooled fast accelerator-driven system has been explored as a thorium-uranium cycle breeder reactor. The sub-critical reactor core operates at an effective neutron multiplication factor of 0.95 and when driven by 1 GeV proton beams of intensity 30 mA, produces about ∼ 900 MWth power. Variation of total thermal power, 233U inventory, Keff, radial and axial power distribution through the operating cycle as well as breeding ratio and doubling time are presented. (author)

  11. Methodical study of cost-benefit analyses of the liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Six American cost-benefit analyses (CBA) of nuclear energy and, in particular, of the Liquid Metal Fast Breeder Reactor (LMFBR) were analysed under the aspect of their methodical difficulties. Two different methodical approaches can be discerned which are related to two completely different applications, according to which the advantages and disadvantages of the breeder reactor are estimated in line with the basic concept of cost-benefit analysis. The analytical methods used to justify the continuation of the breeder-related research programme reveal that the specific energy-related technological and economic conditions of the geographic region considered have to be taken into account. The results of a CBA performed for the USA can therefore not be transferred to the Federal Republic of Germany. Due to the in part strongly differing quantitative results the analyses reviewed do not suggest a clear and final decision in favour of the continuation of the American LMFBR research programme to the extent envisaged. In addition, neither by a positive nor by a negative overall result of the analysis can it be concluded that no other advanced electricity generating technology would have a more favourable cost-benefit ratio, or that the breeder-related research activities, which have been pursued for several years already, should be discontinued. (orig.)

  12. World energy resources, demand and supply of energy, and the prospects for the fast breeder reactor

    International Nuclear Information System (INIS)

    In the past it was taken for granted that the prime role of fast breeder reactors was to complement light water reactors, mainly because of their similar and compatible fuel cycles. In particular, the plutonium converted in LWRs is most intelligently disposed of and used in FBRs. Evaluation of the time horizon of such reactor strategies generally extended only to the year 2000. It is important to realize, however, that the salient task in the breeder field after 2000 - besides electricity generation - will be to substitute for conventional ''cheap'' oil. Electricity today makes up only 10% to 12% of the total secondary energy, while liquids essentially command up to about 50%. Thus the future application of the FBR technology will have to be geared more to the production of a liquid secondary energy carrier than to electricity. A new yardstick for all these considerations is the strongly rising energy prices. They may double, for example, leading to an oil price of US 24/bbl. Under these circumstances it is prudent to generalize the scope for future fast breeders. The key element of such a new fast breeder strategy would be the production of hydrogen by electrolysis or thermolysis or a combination of both. For example, methanol synthesized from hydrogen and residual fossil fuels would thus become economically attractive. The FBR breeding gain, on the other hand, would be used for the continued supply of LWRs generating electricity. The paper identifies order-of-magnitude considerations most important for such a fast breeder application against a global energy demand scenario for the year 2030. (author)

  13. Status and prospects of thermal breeders

    International Nuclear Information System (INIS)

    The main objective of this cooperative study and of this report is to evaluate the extent to which thermal breeders might complement or serve as an alternative to fast breeders in solving the long-term nuclear fuel supply problem. A secondary objective is to consider in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability, and fuel cycle versatility to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  14. Gas core reactors for actinide transmutation and breeder applications. Annual report

    International Nuclear Information System (INIS)

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions

  15. The Last Twenty Years of Experience with Fast Breeder Reactors: Lessons Learnt and Perspectives

    International Nuclear Information System (INIS)

    India has made significant achievements in the design and development of sodium cooled fast breeder reactors over the last twenty years. Attaining a maximum burnup of 165 GW.d/t for the plutonium-rich carbide fuel without any cladding failure, coupled with excellent performance of sodium components, including primary pumps, heat exchangers and steam generators over the last 24 years, reprocessing of carbide fuel with a burnup of 150 GW.d/t and engineering tests performed for validating the plant dynamics computer codes, are the achievements from the successful operation of the 40 MW(th) capacity loop type fast breeder test reactor. Indigenous design of the 500 MW(e) Prototype Fast Breeder Reactor (PFBR), executing high quality multidisciplinary R and D and successful manufacturing and erection of large dimensioned thin walled shell structures are the achievements in PFBR development. These achievements, apart from providing confidence in the PFBR project, are instrumental for the design of innovative future reactors. National and international collaboration established with R and D establishments and academic institutions would go a long way towards helping India to attain world leadership by 2020. (author)

  16. Laboratory analysis of sodium and related materials in the United States Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Keough, R.F.; Ebersole, E.R.

    1978-04-01

    It is essential for the long term operation of a Liquid Metal Fast Breeder Reactor (LMFBR) to control the levels of some sodium and cover gas impurities and to monitor the levels of many others. Chemical analyses are the means of confirming that specific impurities are within established limits. Changes in impurity levels can be most useful in evaluating purification systems and detecting problems such as leaks or excessive corrosion. Extensive analyses are also required with experimental sodium systems related to the Breeder Reactor Program (BRP) since reactor quality sodium is needed to insure relevancy of test data. All of the laboratories use the same procedures for sodium and cover gas analysis. Each year the representatives of the eight laboratories meet to discuss problems they may have and to suggest improvements they would like to see in the methods. Many of these methods are described and their application related to the operation of the Experimental Breeder Reactor II (EBR-II) and Fast Flux Test Facility (FFTF).

  17. Outline of the safety research results, in the power reactor field, fiscal year 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has promoted the safety research in fiscal year of 1996 according to the Fundamental Research on Safety Research (fiscal year 1996 to 2000) prepared on March, 1996. Here is described on the research results in fiscal year 1996, the first year of the 5 years programme, and whole outline of the fundamental research on safety research, on the power reactor field (whole problems on the new nuclear converter and the fast breeder reactor field and problems relating to the power reactor in the safety for earthquake and probability theoretical safety evaluation field). (G.K.)

  18. Outline of the safety research results, in the power reactor field, fiscal year 1996

    International Nuclear Information System (INIS)

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has promoted the safety research in fiscal year of 1996 according to the Fundamental Research on Safety Research (fiscal year 1996 to 2000) prepared on March, 1996. Here is described on the research results in fiscal year 1996, the first year of the 5 years programme, and whole outline of the fundamental research on safety research, on the power reactor field (whole problems on the new nuclear converter and the fast breeder reactor field and problems relating to the power reactor in the safety for earthquake and probability theoretical safety evaluation field). (G.K.)

  19. Status of National Programmes on Fast Breeder Reactors. International Working Group on Fast Reactors, Twentieth Annual Meeting, Vienna, 24-27 March 1987

    International Nuclear Information System (INIS)

    The Agenda of the meeting was as follows: 1. Approval of the Agenda. 2. Approval of the minutes of the 19th meeting of the IWGFR. 3. Report of the Scientific Secretary regarding the WD activities of the Working Group. 4. Presentations and discussions on national programmes on fast breeder reactors. 5. Consideration of conferences on fast breeder reactors. a. ANS-ENS International Conference on Fast Breeder Systems Experience Gained and Path to Economical Power Generation, Richland, Washington, USA, 13-17 September 1987. b. International Conference on Liquid Metal Engineering and Technology, Avignon, France, 17-20 October 1988. c. Other meetings of interest to IWGFR members. 6. Consideration of major recommendations of some of the WD IWGFR Specialists' Meetings. 7. Consideration of arrangements for Specialists' Meetings in 1987. a. Specialists' Meeting on Fission and Corrosion Products Behaviour in Primary Circuits of LMFBRs, Karlsruhe, Fed. Rep. of Germany, May 1987. b. Specialists' Meeting on LMFBR Reactor Block Antiseismic Design and Verification, Bologna, Italy, October 1987. 8. Selection of topics for Specialists' Meetings to be held in 1988 and suggestions of the IWGFR on other Specialists' Meetings and their justifications. 9. Consideration of joint research activities: a. Coordinated Research Programme on a Comparative Assessment of Processing Techniques for Analysis of Sodium Boiling Noise Detection Data. b. Coordinated Research Programme on Intercomparison of LMFBR Core Mechanics Codes. c. New Topics of CRP. d. Other Activities. 10. Updating of ''LMFBR Plant Parameters''. 11. Informal discussion on ''Safety Criteria for Fast Reactors in IWGFR Countries''. 12. The date and place of the 21th Annual Meeting of the IWGFR

  20. Fast breeder reactor reference system classification for the ENEA data bank

    International Nuclear Information System (INIS)

    This report contains the Reference System Classification (RSC) of fast breeder reactors: it provides a functional system breakdown of the reactor. For each system the following important characteristics are reported: the main function, the mode of operation, its location in the reactor, the main interface system, its main components and the component working environment (fluid and/or atmosphere type). The RSC represent a basic step in organizing the ENEA data bank for the registration and processing of reliability data on typical fast reactor components; it provides a functional component breakdown and represent a plant-unique identification in the process of omogenization of event-data coming from different reactors. In this report it was tried to take into account different generations of nuclear power plants, different plant layouts and solutions: in particular loop and pool reactors are separately treated

  1. Safety device for nuclear reactor

    International Nuclear Information System (INIS)

    This invention relates to a safety device for a nuclear reactor, particularly a liquid metal (generally sodium) cooled fast reactor. This safety device includes an absorbing element with a support head connected by a disconnectable connector formed by the armature of an electromagnet at the end of an axially mobile vertical control rod. This connection is so designed that in the event of it becoming disconnected, the absorbing element gravity slides in a passage through the reactor core into an open container

  2. Liquid-metal fast breeder reactor structural materials design considerations

    International Nuclear Information System (INIS)

    This paper gives a brief overview of the LMFBR, to describe its key components, addresses two key structural problems, reviews high-temperature materials utilized, and places bounds on expected operating conditions. The current status of materials utilization in the LMFBR is summarized as follows: with the exception of the reactor upper internals, design needs for the LMFBR can be met with currently approved Code materials; Inconel 718 can potentially solve the thermal striping problems in the reactor upper internals; temperature, stress-strain levels, and design lifetime of the LMFBR push currently approved Code materials toward their limits of usefulness

  3. Applicability of three dimensional diffusion theory programmes based on coarse mesh methods to calculating nuclear characteristics of fast breeder reactors

    International Nuclear Information System (INIS)

    Hexagonal coarse mesh methods in three dimensional diffusion theory programme have been examined for calculating in detail nuclear characteristics of fast breeder reactors composed of hexagonal fuel assemblies, comparing with more accurate triangular fine mesh method. The fast breeder reactors considered here are LMFBRs with different core configurations including heterogeneous core and GCFRs in different burnup states. The nuclear characteristics investigated in the comparative study are effective multiplication factor, power and neutron flux distributions, breeding ratio, reactivity effects and control rod reactivity worth. The comparative study indicates that the conventional coarse mesh method is not adeguate to detailed evaluation on nuclear characteristics of fast breeder reactors, and that the improved coarse mesh method developed by T. Takeda et al. is very useful for this purpose, though some problems exists in evaluation of power distribution and breeding ratio of the extremely composite fast breeder reactors, such as the radially heterogeneous core LMFBR. (author)

  4. In-reactor experiments in fast breeder test reactor for assessment of core structural materials

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled reactor with neutron flux level of the order of 1015 n/cm2/s and temperature of coolant in the range of 650-790K (380-520oC). This reactor is being used as a test bed for the development of fuel and structural materials required for Indian Fast Reactor Programme. FBTR is also used as a test facility to carry out accelerated irradiation tests on thermal reactor structural materials. In-reactor experiments on core structural materials are being carried out by subjecting prefabricated specimens to desired conditions of temperature and neutron fluence levels in FBTR. Non-instrumented irradiation capsules that can be loaded at any location of FBTR core are used for the experiments. Pressurised capsules of zirconium alloys have been developed and subjected to irradiation in FBTR to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy) for life assessment of pressure tubes of Indian Pressurised Heavy Water Reactors (PHWRs). Technology development of pressurised capsules was carried out at IGCAR. These pressurised capsules were filled with argon and a small fraction of helium at a high pressure (5.0-6.5 MPa at room temperature) in such a way that the target stresses were attained in the walls of the pressurised capsules at the desired temperature of irradiation in the reactor. FBTR was operated at a low power of 8 MWt during this irradiation campaign to have an inlet temperature of about 579 K (306oC) which was close to the temperature of pressure tubes at full power in PHWR. Irradiation of thirty pressurised capsules was carried out in FBTR using six irradiation capsules for different durations (upto 79 days). The fluence levels attained by the pressurised capsules were up to 1.1 x 1021 n/cm2 (E> 1 MeV) at temperatures of 579 to 592 K. Post-irradiation increase in diameter of the pressurised

  5. Tritium solubility and permeation in high retention fusion reactor breeder elements

    International Nuclear Information System (INIS)

    As an alternative to the current philosophy of reducing the tritium inventory to a minimum by continuously extracting tritium from the breeder of a fusion reactor, an alternative design philosophy is examined in which tritium is contained within high retention breeder elements which can remain in the reactor for a substantial time. To prevent tritium diffusion through the clad of the element it is necessary to maintain a low tritium pressure within the element. Pressures of between 104 Pa and 1 Pa appear possible with an element containing a high solubility material provided it is kept below about 4000C. This should lead to a leakage into the coolant of between 10 Ci day-1 and 104 Ci day-1 which is considerably less than the 107 Ci day-1 in present designs. (author)

  6. Liquid-metal pumps for large-scale breeder-reactor plant (prototype pump)

    Energy Technology Data Exchange (ETDEWEB)

    Lindsay, M. (comp.)

    1976-07-01

    This report presents the recommended pump design for use in Large Scale Liquid Metal Fast Breeder Reactor plants. The base design for the pump will circulate 127,000 GPM of liquid sodium at temperatures up to 850/sup 0/F and with a total discharge head at the design point of 500 feet Na with an impeller that is 40 feet below the sodium seal. The pump design is predicated on developing an impeller design which will have a suction specific speed (S/sub n/) of about 20,000 with 20 feet NPSH available, which will result in a pump speed of 530 RPM at design conditions. The design is based on the technology developed in the design and fabrication of FFTF pumps, the design efforts for the Clinch River Breeder Reactor Pump design study and other technology.

  7. Tube sheet structural analysis of intermediate heat exchanger for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    The Prototype Fast Breeder Reactor 'Monju' is the first power generating fast breeder reactor in Japan. We have been designing the components of the plant for manufacturing. Among these is the intermediate heat exchanger (IHX) which exchanges heat between primary and secondary sodium loop. The tube sheet of IHX (shell to ligament junction) is a difficult area from the view point of structural strength design under elevated temperature. To validate the structural integrity of tube sheet we performed the series of inelastic analysis and tube sheet thermal shock test using test pieces and half scale model of actual design. The results of inelastic analyses showed there is little progressive deformation around shell to ligament structural discontinuous junction. Furthermore, thermal shock tests showed no increase of an accumulative deformation. By these analyses and experiments, structural reliability of tube sheet could be shown. (author)

  8. Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

    International Nuclear Information System (INIS)

    The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

  9. Status of the fast breeder reactor technology in China

    International Nuclear Information System (INIS)

    According to the Chinese long-term energy strategy the FBR development is strongly supported. In the near term nuclear programme it is intended to build the experimental First Fast Reactor (FFR) in the year 2000. Design work is in progress. (author). 1 ref., 6 figs, 8 tabs

  10. Present status and problems of development of fast breeder reactors

    International Nuclear Information System (INIS)

    The development of FBRs in Japan now reached the stage to conclude on the development organization for a demonstration reactor positioning one step before a practical reactor. FBRs can be operated while converting uranium-238 existing in natural uranium by 99.3% to fissile plutonium-239, as the result, the nuclear fuel more than that consumed can be produced. However, there are various technical difficulties in FBRs, and the construction cost is estimated to be considerably higher as compared with that of LWRs. Also the plutonium obtained by reprocessing spent fuel is used for FBRs, accordingly, the development of FBRs is inseparable from the establishment of nuclear fuel cycle. In order to get rid of the burden of enormous development cost for FBRs, the trend of international joint development is conspicuous. The Superphenix with 1200 MWe output under construction centering around SERENA is expected to attain the criticality in the spring of 1985. For the development of a demonstration reactor, it is necessary to increase the role of private businesses, and the smooth transfer of know-how accumulated in Power Reactor and Nuclear Fuel Development Corp. to civilian side is an important problem. (Kako, I.)

  11. Application of microprocessor based controller in the Breeder Reactor Program

    International Nuclear Information System (INIS)

    This paper treats Argonne National Laboratory's experience using microprocessor based controllers presently in use on several control loops within the EBR-II reactor facility as well as tests being performed by these controllers. Also included is a discussion of the expandability, modularity, range of capabilities and higher level functions possible using such equipment

  12. Fast Breeder Reactor Development in France During 1987

    International Nuclear Information System (INIS)

    On March 8, 1987, a ''sodium leak'' alarm signal was received in the Creys-Malville control room. By the end of March, it had been established beyond all doubt that sodium was, in fact, leaking into the fuel storage drum inter-vessel gap. The reactor has been shut down since May 26. The origin of the leak was located on September 5, after complete drainage of the main tank. Despite the fact that the leak was confined, had had no radiological consequences and cast no doubts on the safe operation of the reactor, the impact of this incident on public opinion, both in France and in the neighbouring European states, was considerable. Two facts would appear to have been decisive. The first was that the reactor had not been shut down immediately, the second was that the leak was only detected and localized in September: it was difficult for people to understand that before its exact position could be determined, certain operations (transfer of a few subassemblies to the reactor core, unloading of the fuel storage drum) had to be performed

  13. Analysis of principle possibilities of intermediare storage of fast breeder reactor fuel elements

    International Nuclear Information System (INIS)

    The principle possibilities of intermediate storage of fast breeder reactor fuel elements were analyzed and compared on the basis of 4 different concepts of storage. The SNR-2 fuel element was chosen as reference. Only the pool (wet) storage could be used to store fuel elements of less than 18 months precooling time. The other concepts (dry storage and container storage) have distinct advantages at precooling times longer than 18 months. (orig./HP) With 22 tabs., 8 figs

  14. Seismic parametric studies in a large scale prototype breeder reactor plant

    International Nuclear Information System (INIS)

    Seismic parametric studies were conducted for a large scale prototype breeder reactor plant (135C MW). The effects of plant configuration, soil stiffness and deep embedment were evaluated. Nuclear island interconnected structures on a common foundation mat with a symmetrical arrangement resulted in lower seismic responses. All other conditions being equal, soft sites are preferable to stiff sites. Deep embedment of the nuclear island could, in certain sites, result in a reduction of seismic responses. (orig.)

  15. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    International Nuclear Information System (INIS)

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented

  16. Accident consequence studies for large fast breeder reactor containments built of concrete or steel

    International Nuclear Information System (INIS)

    A numerical analysis of accident consequences in a fast breeder reactor of commercial size after complete loss-of-heat-sink was performed, using the CONTAIN code. Two containment types were studied, which differ in the material used for shielding, support and confinement structures. It was found that the replacement of concrete as principal construction material by steel offers a significant potential for consequence mitigation in terms of thermal and pressure loads and of retention capability

  17. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  18. Present day design challenges exemplified by the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    The present day design challenges faced by the Clinch River Breeder Reactor Plant engineer result from two causes. The first cause is aspiration to achieve a design that will operate at conditions which are desirable for future LMFBRs in order for them to achieve low power costs and good breeding. The second cause is the licensing impact. Although licensing the CRBRP won't eliminate future licensing effort, many licensing questions will have been resolved and precedents set for the future LMFBR industry

  19. Application of mass-predictions to isotope-abundances in breeder-reactor cores

    International Nuclear Information System (INIS)

    The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGRN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technically view-point. (orig.)

  20. Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

    International Nuclear Information System (INIS)

    This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser

  1. Application of mass-predictions to isotope-abundances in breeder-reactor cores

    CERN Document Server

    Kirchner, G

    1981-01-01

    The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGEN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technical viewpoint. (15 refs).

  2. Research and developments on nondestructive testing in fabrications of fast breeder reactor structural components in Japan

    International Nuclear Information System (INIS)

    Research and developments (R and D) have been conducted on the nondestructive testing techniques necessary for the construction of fast breeder reactor (FBR). Radiographic tests have been made on tube-tube plate welds and small-diameter tube welds, etc. Ultrasonic tests have been conducted on austenitic stainless steel welds. In the penetrant tests and magnetic particle tests, the investigations have been performed on the effects of various test factors on the test results

  3. Fast breeder reactor. The past, the present and the future. (7) History of fast reactor development in Japan - 2

    International Nuclear Information System (INIS)

    History and present state of fast breeder reactor was reviewed in series. As a history of fast reactor development in Japan - 2, this seventh lecture presented the development of the prototype FBR (MONJU) and design studies of the demonstration reactor. The MONJU started operation in 1994, but a sodium leakage in its secondary heat transfer system occurred during performance tests in 1995. It has not operated since and activities for restart are conducted. Since 1997 design studies of the demonstration FBR have been conducted to reflect the MONJU sodium leakage accident and also establish its economic competitiveness with advanced LWR. (T. Tanaka)

  4. Status of national programmes on fast breeder reactors. Nineteenth annual meeting, Kalpakkam, India, 11-14 March 1986

    International Nuclear Information System (INIS)

    The Nineteenth Annual Meeting on the Status of National Programmes in Member States of the IAEA on Fast Breeder Reactors had been held in March 1986. The representatives of the Member States and international organizations reported status and activities in the field of fast breeder reactors development and operation. A report on uranium supply and demand was also presented by the NEA/OECD. A separate abstract was prepared for each of the 11 presentations of the meeting

  5. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

  6. An estimate of the radiological consequence of notional accidental releases of radioactivity from a fast breeder reactor

    International Nuclear Information System (INIS)

    In this report an assessment is made of the radiological consequences of notional accidental releases of activity from a fast breeder reactor under certain circumstances. It was prepared under contract to the Nuclear Installations Inspectorate (Health and Safety Executive) to assist them in making a preliminary safety assessment of fast reactors. The range of releases considered in the report was specified by the Inspectorate and comprises the vaporisation and release of varying fractions of the core of a 1300 MW(e) reactor. Two cases are evaluated depending on assumptions relating to the remainder of the core. No attempt is made to assign any probability to the occurrence of a given release; the report provides no more than a part of the information necessary for a safety assessment and is to be considered only within this limited context. The subject is dealt with under the following headings: introduction; parameters used in the assessment; atmospheric dispersion; pathways of exposure and dosimetric models; doses associated with the release of one tonne of fuel; biological effects; consequences of releases in which the nuclide composition differs from that in the fuel. (U.K.)

  7. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    International Nuclear Information System (INIS)

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 25800F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

  8. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    International Nuclear Information System (INIS)

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  9. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev, E-mail: baldev.dr@gmail.com

    2015-09-15

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  10. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    Science.gov (United States)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  11. Physics calculations for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah; Kier, P.H.; Hummel, H.H.

    1977-06-01

    Calculations of distributions of power and sodium void reactivity, unvoided and voided Doppler coefficients and steel and fuel worths have been performed using diffusion theory and first-order perturbation theory for the LWR discharge Pu-fueled CRBR at BOL, the FFTF-grade Pu-fueled CRBR at BOL and for the beginning and end of equilibrium cycle of the LWR-Pu-fueled CRBR. The results of the burnup and breeding ratio calculations performed for obtaining the reactor compositions during the equilibrium cycle are also reported. Effects of sodium and steel contents on the distributions of sodium void reactivity and steel worth have also been studied. Errors and uncertainties in the reactivity coefficients due to cross-sections and the two-dimensional geometric representations of the reactor used in the calculations have also been estimated. Comparisons of the results with those in the CRBR PSAR are also discussed.

  12. Contained fissionly vaporized imploded fission explosive breeder reactor

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

  13. Fast-core thermal-blanket breeder reactor

    International Nuclear Information System (INIS)

    A preliminary assessment of the performance expected from a specific type of FCTB reactor, consisting of a gas-cooled fast system for the core and natural-uranium light-water thermal system for the blanket is reported. Both the core and the blanket use the 238U-Pu fuel cycle. When all the neutrons leaking out of the core reach the blanket, the blanket-to-core power ratio is estimated to be about 1.3. By reducing its water-to-fuel volume ratio below 1.5, the light water blanket can be designed to have a higher ksub(eff), while maintaining an equilibrium fissile fuel content. Compared with conventional FBRs, having the same power output, the FCTB reactor considered offers the following advantages: a lower fissile fuel content, easier and safer control, no need for Pu separation. (B.G.)

  14. Conceptual design of a pool type molten salt breeder reactor

    International Nuclear Information System (INIS)

    The renewed interest in molten salt coolant technology is backed by the 50 years history of molten salt nuclear technology development, mainly in Oak Ridge National Laboratory (ORNL). In Indian context MSBR is found to be one of the options for sustainable nuclear energy generation, especially in the third stage of the nuclear programme. The system can be operated at high temperature which makes high efficiency power conversion and efficient hydrogen generation through thermo-chemical reactions possible. At present development is in progress in BARC on two molten salt reactor concepts, one is pool type and the other is loop type. Here the design of pool type concept with 850MWe power is described. The core is designed to operate in the fast spectrum region so the conversion of 233U breeding is possible from thorium. Preliminary thermal hydraulic analysis is carried out with LiF-ThF4-UF4 as the primary fuel and coolant. The blanket material is also a molten salt, LiF-ThF4. Reactor physics calculations are also carried out for the feasibility studies of the core design of the reactor. FLiNaK is used as the secondary coolant for the calculations. Both forced circulation and natural circulation options are evaluated. (author)

  15. Conceptual design of loop-in-tank type Indian molten salt breeder reactor concept

    International Nuclear Information System (INIS)

    The third stage of Indian nuclear power programme envisages use of thorium as fertile material with 233U, which is proposed to be obtained from reprocessing of spent fuel of Pu/Th based fast reactors in the later part of the second stage of the programme. In India, thorium based reactors have been designed in many configurations, from light water cooled designs to high temperature liquid metal and molten salt cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). (author)

  16. The fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U3O8). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-pathitem for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices

  17. Status of national programmes on fast breeder reactors. Twenty-fifth annual meeting of the International Working Group on Fast Reactors. Summary report. Working material

    International Nuclear Information System (INIS)

    At present nuclear power accounts for approximately 17% of total electricity generation worldwide. Given continuing population growth and the needs of the third world and developing countries to improve their economic performance and standard of living, energy demand is expected to continue to grow through the 21st century. The proportion of energy supplied as electricity is also expected to continue to increase. Although fossil fuelled electricity generation is the option preferred by several countries for the short term, there are rising concerns over climatic consequences caused by extended burning of fossil fuels as a result of the demands of a fast expanding world population. In this situation nuclear electricity will become more and more important and the known reserves of uranium would be consumed quite quickly by thermal reactors. It would be possible to sustain a large nuclear programme only by introducing fast reactors. One can conclude that there are strategic reasons for pursuing the development of fast breeder reactors. It will become desirable essential, to have this technology available for introduction. The experience of the various prototypes presently in operation has confirmed the operability and benign characteristics of the LMFR and has given ground for confidence in the future. Current fast reactor designs offer very large margins of safety and by virtue of redundant and diverse safety systems the potential for an energetic core disruptive accident or for fast reactor core meltdown has been essentially eliminated. Several international forums reviewed the current trends in the fast reactor development. The view was reaffirmed that fast breeder reactors still remain the most practical tool for effective utilization of uranium resources for the future energy needs. Achievement of competitiveness with LMRs is still the first priority condition for the future deployment of this type of reactor. The recycling of plutonium into LMFBRs would allow

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

    International Nuclear Information System (INIS)

    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

  19. Studies of the restructuring of fast breeder test reactor fuel by out-of-pile simulation

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) at Kalpakkam, India, currently employs a mixed carbide of uranium and plutonium with a Pu/(Pu + U) ratio of 0.70 as fuel. The behavior of this fuel in a thermal gradient is investigated. An out-of-pile simulation facility is designed, set up, and commissioned. Experiments are conducted on FBTR fuel pellets to study the restructuring of the fuel at various levels of linear power and its cracking behavior in a thermal gradient. The results are discussed in terms of their significance for reactor operation

  20. Civilian nuclear power on the drawing board: the development of Experimental Breeder Reactor-II

    International Nuclear Information System (INIS)

    On September 28, 2001 a symposium was held at Argonne National Laboratory as part of the festivities to mark the 100th birthday of Enrico Fermi. The symposium celebrated Fermi's ''contribution to the development of nuclear power'' and focused on one particular ''line of development'' resulting from Fermi's interest in power reactors: Argonne's fast reactor program. Symposium participants made many references to the ways in which the program was linked to Fermi, who led the team which created the world's first self-sustaining nuclear chain reaction. For example, one presentation featured an April, 1944 memo that described a meeting attended by Fermi and others. The memo came from the time when research on plutonium and the nuclear chain reaction at Chicago's WWII Metallurgical Laboratory was nearing its end. Even as other parts of the Manhattan Engineering Project were building on this effort to create the bombs that would end the war, Fermi and his colleagues were taking the first steps to plan the use of nuclear energy in the postwar era. After noting that Fermi ''viewed the use of [nuclear] power for the heating of cities with sympathy,'' the group outlined several power reactor designs. In the course of discussion, Fermi and his colleagues took the first steps in conjuring the vision that would later be brought to life with Experimental Breeder Reactor I (EBR-I) and Experimental Breeder Reactor II (EBR-II), the celebrated achievements of the Argonne fast reactor program. Group members considered various schemes for a breeder reactor in which the relatively abundant U-238 would be placed near a core of fissionable material. The reactor would be a fast reactor; that is, neutrons would not be moderated, as were most wartime reactors. Thus, the large number of neutrons emitted in fast neutron fission would hit the U-238 and create ''extra'' fissionable material, that is, more than ''invested,'' and at the same time produce power. The group identified the problem of

  1. Civilian nuclear power on the drawing board: the development of Experimental Breeder Reactor-II.

    Energy Technology Data Exchange (ETDEWEB)

    Westfall, C.

    2003-02-20

    On September 28, 2001 a symposium was held at Argonne National Laboratory as part of the festivities to mark the 100th birthday of Enrico Fermi. The symposium celebrated Fermi's ''contribution to the development of nuclear power'' and focused on one particular ''line of development'' resulting from Fermi's interest in power reactors: Argonne's fast reactor program. Symposium participants made many references to the ways in which the program was linked to Fermi, who led the team which created the world's first self-sustaining nuclear chain reaction. For example, one presentation featured an April, 1944 memo that described a meeting attended by Fermi and others. The memo came from the time when research on plutonium and the nuclear chain reaction at Chicago's WWII Metallurgical Laboratory was nearing its end. Even as other parts of the Manhattan Engineering Project were building on this effort to create the bombs that would end the war, Fermi and his colleagues were taking the first steps to plan the use of nuclear energy in the postwar era. After noting that Fermi ''viewed the use of [nuclear] power for the heating of cities with sympathy,'' the group outlined several power reactor designs. In the course of discussion, Fermi and his colleagues took the first steps in conjuring the vision that would later be brought to life with Experimental Breeder Reactor I (EBR-I) and Experimental Breeder Reactor II (EBR-II), the celebrated achievements of the Argonne fast reactor program. Group members considered various schemes for a breeder reactor in which the relatively abundant U-238 would be placed near a core of fissionable material. The reactor would be a fast reactor; that is, neutrons would not be moderated, as were most wartime reactors. Thus, the large number of neutrons emitted in fast neutron fission would hit the U-238 and create ''extra'' fissionable material

  2. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    The results of activities carried out by the DeBeNe partners in 1988 have been compiled in this report. The 1989 KNK II experimental program will focus on the management of fuel element failures. This includes in particular post-irradiation examinations in the hot cells and the THIBO experiments (thermally induced fuel rod oscillation). For this program, nine permits were received in 1988 for the installation and operation of test systems, including a new facility for localizing failed fuel elements. Increasing the purity of sodium is the purpose of a cesium trap recently installed, and of modifications to an oxygen probe operated for test purposes. The SNR 300 project is being continued within the framework of the ''holding phase''. The objective of this phase of operation is to keep the reactor plant in the loading and operational states and execute planning within the licensing procedure necessary to obtain the next partial permit. R and D work was concentrated on fuel and materials development, safety, physics, and components development. Materials for fuel, blanket, and absorber elements were studied and further developed with a view to very high burnup. In the sector of physics, the engineering design and the nuclear design of large breeders call for a reduction of the margins of uncertainty in reliable predictions of the major reactor parameters. The development of the joint European cell code, ECCO (European Cell Code) has progressed far enough in the meantime to allow the criticality parameter, keff, of a cell arrangement to be calculated with ECCO for the first time at UKAEA Winfrith, the central agency for carrying out the development work. One of the major aspects covered in safety studies is the complex of fuel rod failures, loss of coolant flow, and power transients as possible causes of accidents. Studies conducted into the management of credible accident consequences were concentrated, among other topics, on the behaviour of aerosols, sodium fires

  3. Programme and status of the development of the fast breeder reactor system in France

    International Nuclear Information System (INIS)

    The fundamental justification for fast breeder reactors is based on their breeding ability, and the development of this reactor line is a matter of national importance because there is no reasonable and adequate alternative that could both cover the increasing future energy requirements and provide energy independence for France. The 250MW(e) Phenix power plant has been satisfactorily operated as a demonstration plant for two years now. This proof of the validity of the system was necessary. It underlined its reliability and confirmed the value of the process to the constructors and to the domestic utility, EDF, who confidently decided on industrial and commercial commitment to the new reactor line. This attitude is demonstrated in the following actions in France: (a) the construction by Nersa of the Super-Phenix power plant, a 1200MW(e) prototype commercial power plant which will provide experience in building and operating large fast-neutron power plants; (b) the firm intention of EDF to proceed and bring breeders into operation in the 1990s, representing about 8000MW(e); (c) the creation of a strong industrial organization: the creation in France of the Novatome Company corresponds to this objective - Novatome is now responsible for the industrial development of fast breeder reactors with Creusot-Loire as industrial basis and with all the know-how owned by CEA; (d) the Compagnie generale des matieres nucleaires, Cogema, a CEA subsidiary, is reviewing the actions necessary to reach industrial levels for the cost of the whole fuel cycle. On the other hand, the negotiation of international agreements adds the knowledge and experience gained in the R and D programmes of the various countries concerned. They improve the value and the technical, industrial and commercial basis of this new type of reactor, thus enlarging the prospects of development and commercialization. (author)

  4. Cladding and wrapper development for fast breeder reactor high performance

    International Nuclear Information System (INIS)

    In order to ensure economic performance, of both the existing reactors and the future EFR, much recent research has been carried out within the framework of the European R and D agreement to examine the properties of various wrapper and cladding alloys. This paper reviews the status of the European research and development programmes on these steels and highlights the most striking results. For the cladding alloys, results on dimensional stability and tensile properties for fuel pin cladding irradiated in PFR or Phenix will be given. As for wrappers the presently available results of those wrappers irradiated in Phenix and PFR show that both ferritic steels are very good candidates and that on the basis of our present knowledge most of the properties are satisfactory for wrapper applications

  5. Liquid Metal Fast Breeder Reactor Program (LMFBR): facility profiles

    International Nuclear Information System (INIS)

    A description is presented of the experimental test facilities involved in the conduct of the LMFBR research and development program. Existing facilities and those under construction or authorized as of October 1975 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. Introductory material for each section includes site and facility maps and an alphabetical list of the profiles contained in the section. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc. involved in the LMFBR program is included. Alphabetical, organizational, and programmatic indexes are provided as a convenient method of identifying the facilities with their locations and with their principal uses in the LMFBR program

  6. Experience in the maintenance of sodium systems of fast breeder test reactor

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam in India and that has been operating for 25 years. The reactor has been operated up to a power level of 18.6 MWt with a sodium outlet temperature of 482 C. degrees. Several modifications were carried out in the sodium systems to improve the plant performance. During the course of operation of the reactor, a number of sodium laden components like pumps, valves, cold traps, rupture disks, level probes, shielding plugs, control rod drive mechanisms, experimental assemblies, piping... were removed for various maintenance, modification and replacement jobs which has given the operators a valuable experience in handling large scale sodium systems. This paper details the special procedures followed during the handling of active and inactive sodium laden components

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

    International Nuclear Information System (INIS)

    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

  8. Development of metallic fuels for Indian Fast Breeder Reactors

    International Nuclear Information System (INIS)

    The neutronic performance of metal fuel based on binary U-Pu alloy or ternary U-Pu-Zr alloys are better than conventional uranium plutonium mixed oxide or high density carbide ceramic fuel. The growing energy demand in India needs faster growth of nuclear power and warrants introduction of fast reactors based on metallic fuels in future. Physics calculation showed that fast reactor based on metallic fuels results in higher breeding ratio and lower doubling time compare to mixed oxide or carbide fuels. Moreover inclusion of pyro-processing of the fuel in the fuel cycle is expected to make metal fuel option more economical. As part of metal fuel development programme for future FBR's in India, capsule irradiation of metal fuel based on sodium bonded U-Pu-Zr alloy and metal (Zircaloy) bonded binary U-Pu (Pu ∼ 15 %) alloy are being actively pursued. For this purpose two design concepts have been proposed : one based on sodium bonded ternary alloy fuel of U-Pu-Zr (2-10 wt%) in modified T91 cladding material and the other is U-Pu binary alloy mechanically bonded to modified T91 cladding material with 'Zircaloy' as a liner between the fuel alloy and the clad. The Zircaloy liner act as a barrier in reducing the fuel clad chemical interaction. It also helps in transfer of heat from the fuel to the clad. The smear density of metal bonded pin will be between 70% - 85% and that for sodium bonded pin will be ∼ 70%. In metal bonded fuel pin design two/four semi-circular grooves of diameter ∼1.0 mm, will be provided in diametrically opposite directions in the fuel cross section to accommodate fuel swelling. A comparison has been made on the relative merits and demerits of these two fuel pin designs. The material for the axial blanket will be 'U' or U-Zr (Zr upto 10wt %) alloy based on the results of the out-of-pile thermal cycling behavior and irradiation performance. In the present investigation out-of-pile experiments have been carried out to address some of the issues of

  9. Meeting on reactor safety research

    International Nuclear Information System (INIS)

    The meeting 'Reactor Safety Research' organized for the second time by the GRS by order of the BMFT gave a review of research activities on the safety of light water reactors in the Federal Repulbic of Germany, international co-operation in this field and latest results of this research institution. The central fields of interest were subjects of man/machine-interaction, operational reliability accident sequences, and risk. (orig.)

  10. Reactor safety. A literature search

    International Nuclear Information System (INIS)

    Included are 677 citations to report literature abstracted in Nuclear Abstracts during the period July 1973 through December 1974 on safety aspects of nuclear reactor design, siting, materials, engineered safeguards;and operating procedures. The reports cited concern the reliability of reactor systems and components, the consequences of potential accident conditions, and safety in power plant operations. Subject, author, and report number-availability indexes are included. (U.S.)

  11. Reactor safety: a literature search

    International Nuclear Information System (INIS)

    Included are 905 citations to report literature abstracted in Energy Research Abstracts during the period July 1976 through May 1977 on safety aspects of nuclear reactor design, siting, materials, engineered safeguards, and operating procedures. The reports cited concern the reliability of reactor systems and components, the consequences of potential accident conditions, and safety in power plant operations. Subject, author, corporate author, and report number-availability indexes are included

  12. Fast breeder reactor. The past, the present and the future. (6) History of fast reactor development in Japan - 1

    International Nuclear Information System (INIS)

    History and present state of fast breeder reactor was reviewed in series. As a history of fast reactor development in Japan - 1, this sixth lecture presented the start of FBR development, and construction and operation of the experimental FBR (JOYO). The JOYO began operation in 1977 and now is being operated at 140 MWt after two times of upgraded modification. The JOYO is aimed at (1) advancement of technology through and experiment, (2) conducting irradiation tests on fuels and materials and (3) validation of innovative technology for development of a future FBR. (T. Tanaka)

  13. Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

  14. Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

    International Nuclear Information System (INIS)

    Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component

  15. Sodium and steam generator leak detection for prototype fast breeder reactor (PFBR)

    International Nuclear Information System (INIS)

    The construction of the Prototype Fast Breeder Reactor (PFBR) a 500 MWe pool type sodium cooled breeder reactor with MOX fuel has started at Kalpakkam. The Instrumentation and Control of PFBR is designed for safe, reliable and economic operation of the plant. Special feature of breeder reactors is sodium instrumentation. Leaks in sodium systems have the possibility of being exceptionally hazardous due to the reaction of liquid sodium with oxygen and water vapour in the air. In addition, leakage from primary systems can cause radioactive contamination. Potential regions of leakage are near welds and high stress areas. Sodium also reacts with concrete releasing hydrogen and leading to damage and loss of strength of concrete structures. Leaking sodium catches fire depending on its temperature. Sodium temperature in the plant ranges from 423 K at filling condition to 820 K at reactor nominal power operating condition. Leak detectors are provided on pipelines, tanks and other capacities. Sodium leak detection systems are designed to meet requirements of ASME section XI- division 3 which specifies that sodium leak at the rate of 100 g/h are to be detected in 20 h for air filled vaults and 250 h for inert vaults. Diverse leak detection methods are employed for active and non-active sodium equipment and pipes. For detection of water leaks into Sodium in steam generators, Hydrogen in Sodium Detectors (HSD) are used. Hydrogen in Argon Detectors (HAD) are used for sodium temperatures below 623 K as HSD is not effective below this temperature due to non-dissolution of hydrogen formed. Choice and challenges posed in implementation of above leak detection requirements are discussed in this paper. (authors)

  16. Light water reactor safety research

    International Nuclear Information System (INIS)

    As the technology of light water reactors (LWR) was being commercialized, the German Federal Government funded the reactor safety research program, which was conducted by national research centers, universities, and industry, and which led to the establishment, in early 1972, of the Nuclear Safety Project in Karlsruhe. In the seventies, the PNS project mainly studied the loss-of-coolant accident. Numerous experiments were run and computer codes developed for this purpose. In the eighties, the Karlsruhe Nuclear Research Center contributed to the German Risk Study, investigating especially core meltdown accidents under the impact of the events at Three Mile Island-2 and Chernobyl-4. Safety research in the nineties is concentrated on the requirements of future reactor generations, such as the European Pressurized Water Reactor (EPR) or potential approaches which, at the present time, are discernible only as tentative theoretical designs. (orig.)

  17. Development of an ISI robot for the fast breeder reactor MONJU primary heat transfer system piping

    International Nuclear Information System (INIS)

    The fast breeder reactor (FBR) 'MONJU' carry out in-service inspection (ISI) in important components for safety. ISI of the primary heat transfer system (PHTS) piping is performed by sodium leak monitoring, a visual testing with ITV camera and a volumetric testing with ultrasonic. The volumetric testing inspect maximum part of stress concentration in PHTS pipe by using ultrasonic. ISI use remote control robot on the grounds of high temperature (atmosphere 55 deg. C, pipe surface 80 deg. C) and radiation exposure condition (dose rate 10mGy/h, pipe surface dose rate 15mGy/h). Moreover, volumetric testing use tire type ultrasonic sensor on the grounds of a sodium boundary which chemically reacts with water and oil. Light-water reactors (LWR) can be inspected by ultrasonic that uses water and oil. Purpose. This development of inspection system is intended to use new control robot and new tire type ultrasonic sensor. The robot control adopt teaching control method. The target is reproducibility of less than ±5mm. The new tire type ultrasonic sensor adopt double oscillators, because of the multipath reflection wave from contact rubber etc., the noise level decreases and consequently S/N ratio well. The defective detection target was decided to be a depth 50% electrical discharge machining (EDM) slit from pipe wall thickness (t=11.1mm) with a signal per noise ratio (S/N) not less than 2 (6dB). Results and Conclusions We developed a new inspection system for the in-service inspection of PHTS of the FBR 'MONJU'. Moreover, we carried out performance test about new inspection system. The control performance of the new robot driving confirmed it was about less than 5mm by the experiment. The detection performance of new tire sensor confirmed it was detectable an EDM slit with depth 10% from pipe thickness and with a S/N ratio not less than 4.0 (12.0dB). The robot and new tire sensor that developed as a result of the experiment confirmed the performance that was able to be

  18. Social and ethical aspects of the liquid-metal fast breeder reactor

    International Nuclear Information System (INIS)

    Development of liquid fast breeder reactors not only indirectly entails (through commitments of time and resources that foreclose other options), but also directly entails large-scale centralized electrification. The massive economic commitments of such a policy, wether or not it is a nuclear policy, demand and cause major social changes, bypass traditional market mechanisms, concentrate political and economic power, persistently distort political structures and social priorities, compromise professional ethics, are probably inimical to greater distributional equity within and among nations, enhance vulnerability and the paramilitarization of civilian life, introduce major economic and social risks, and reinforce current trends toward centrifugal politics. Deployment of fission technology produces further social and ethical problems, since attempts to reduce potential hazards from operating accidents, from escape of nuclear wastes, or from nuclear violence and coercion will have socio-political side-effects even if they succeed, not to mention the side-effects if they fail. These side-effects, many of which would be worse with fast than with thermal reactors, include repressiveness, abrogation of civil liberties, social rigidity and homogeneity, elitist technocracy, dirigiste autarchy, and suppression of ethical objections. The inability of modern political institutions to cope with the persistent hazards of toxic and explosive nuclear materials strains the competence and perceived legitimacy of those institutions as they try to compromise between individual liberties and public safety and to subject to democratic decision technically tinged policy questions that turn largely on unknown or unknowable information. There is no scientific basis for calculating the likelihood on the maximum long-term of nuclear mishaps, nor for guaranteeing that the effects will not exceed a particular level; it is only known that all precautions are, for fundamental reasons

  19. Thermochemical study of material compatibility for sodium cooled fast breeder reactor application

    International Nuclear Information System (INIS)

    In fast breeder reactors, liquid sodium is preferred as a coolant due to its high thermal conductivity, high specific heat, low viscosity, wide liquid range, remarkable thermal stability. However, it must be in the pure form to be compatible with structural materials in which chemical compounds of Austenitic stainless used include carbon, chromium, nickel, molybdenum, iron, niobium, zirconium and so on. Traces of impurities play an important role in corrosion, mass transport loops of the reactor. Corrosion of structural materials in liquid sodium is deeply affected by the oxygen concentration. Some of these corrosion products which find their way into sodium can cause risk when they deposit on parts like heat exchangers and pumps, which have to be periodically maintained. Thus one must not only control and monitor the oxygen impurity level, but also understand the mechanism of the chemical reaction in the reactor. In this way, thermodynamic approach is obtained by analyzing compatibility of chemical compounds of structural materials with liquid sodium

  20. The fast breeder reactor Rapsodie (1962); Le reacteur rapide surregenerateur rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L.; Zaleski, C.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [French] Dans ce rapport, les auteurs font le point du projet RAPSODIE (reacteur francais surregenerateur a neutrons rapides), au moment du debut effectif de sa construction. On y trouvera decrits: les principales caracteristiques neutroniques et thermiques, le bloc pile et les circuits de refroidissement, les principaux moyens de manutention des ensembles actifs ou contamines, les principes et les moyens qui regissent la conduite du reacteur, les fonctions et l'implantation des divers batiments. La divergence de RAPSODIE est prevue pour 1964. (auteurs)

  1. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  2. Thermal reactor safety

    International Nuclear Information System (INIS)

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport

  3. Thermal reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  4. Clinch River Breeder Reactor Plant steam generator: FEW tube test model post test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) is part of an extensive testing program being carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. The problems encountered with the mechanical features of the FTT model design which led to premature test termination and the results of the post-test examination are described

  5. A supercritical steam cooled fast breeder reactor with negative reactivity characteristics against voiding and flooding

    International Nuclear Information System (INIS)

    The steam cooled fast breeder reactor with negative reactivity characteristic against voiding and flooding is feasible under the supercritical pressure. The breeding ratio is 1.04. A flat core with the zirconium hydride layer is adopted for mitigating the void reactivity. The thermal efficiency of the indirect cycle system is improved 9 % relatively from the current PWR's. The core should be cooled in 10 seconds after the large break loss of coolant accident (LOCA). The coast down time should be larger than 30 seconds to overcome the loss of flow (LOF) by the trip of all blowers. (author)

  6. Description of a materials/coolant laboratory for support of the Breeder Reactor Technology Shipping Program

    International Nuclear Information System (INIS)

    A description of a facility devoted to evaluating the environmental compatibility and mechanical response of materials suitable for a breeder reactor spent-fuel shipping cask is given. The facility presently consists of a closed-loop servo-controlled hydraulic, horizontal test system coupled to an environmental chamber, generalized mechanical test equipment and high-rate mechanical behavior apparatus. Future plans include the procurement of real-time computer control equipment which will be used to assess the effects of complex load-time histories on spent-fuel shipping cask materials

  7. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This final report is a complete conceptual design study of a mechanical pump for a large scale breeder reactor plant. The pumps are located in the cold leg side of the loops. This makes the net positive suction head available - NPSHA - low, and is, in fact, a major influencing factor in the design. Where possible, experience gained from the Clinch River Project and the FFTF is used in this study. Experience gained in the design, manufacturer, and testing of pumps in general and sodium pumps in particular is reflected in this report. The report includes estimated cost and time schedule for design, manufacture, and testing. It also includes a recommendation for development needs.

  8. Comparison of diffusion and transport theory analysis with experimental results in fast breeder test reactor

    International Nuclear Information System (INIS)

    A systematic analysis has been performed by 3 dimensional diffusion and transport methods to calculate the measured control rod worths and subassembly wise power distribution in fast breeder test reactor. Geometry corrections (rectangular to hexagonal and diffusion to transport corrections are estimated for multiplication factors and control rod worths. Calculated control rod worths by diffusion and transport theory are nearly the same and 10% above measured values. Power distribution in the core periphery is over predicted (15%) by diffusion theory. But, this over prediction reduces to 8% by use of the SN method. (authors). 9 refs., 4 tabs., 3 fig

  9. Level-2 PSA for the prototype fast breeder reactor MONJU applied to the accident management review

    International Nuclear Information System (INIS)

    An accident management guideline (AMG) of the prototype fast breeder reactor MONJU has been presented to Nuclear and Industry Safety Agency (NISA) of METI by Japan Atomic Energy Agency (JAEA) with an evaluation result of an effectiveness of the AMG by employing Level-1 and Level-2 PSAs. Japan Nuclear Energy Safety Organization (JNES) evaluated the three events - PLOHS, LORL and ATWS events - and scrutinized the results of the Level-2 PSA carried out by JAEA from the view point of an accident management (AM) review. Regarding ATWS events, we have carried out a qualitative evaluation of the results of JAEA's evaluation and carried out a quantitative evaluation of the containment failure frequency (CFF) in relation to Protected-Loss-of-Heat-Sink (PLOHS) and Loss-of-Reactor-Level (LORL) events. Evaluation of the containment failure probability CFF has been conducted based on the results of the Level-1 PSA by employing the code system developed by JNES. We conducted a close examination of the procedure that JAEA followed to evaluate CFFs in PLOHS and LORL events. It was confirmed that JAEA's Level-2 PSA quantified the phenomenal event trees was expanded in the three processes - the plant response process, the core damage process and the containment vessel response process - based on various analytical and experimental evidence and otherwise followed much the same basic evaluation procedures employed by JNES. As for PLOHS and LORL, quantitative evaluation of CFF was conducted according to the following procedures: Development of an event flow diagram, Development of a phenomenal event tree, Quantification of the phenomenal event tree, Evaluation of containment failure frequencies, and Evaluation of the effectiveness of the AM measures. In the evaluation of the PLOHS and LORL events, the following analytical codes were used; Plant dynamic characteristic analytical code (NALAP-II), Nuclear characteristics analytical system (ARCADIAN-FBR/MVP), Nuclear dynamics analysis code

  10. Optimization of the Westinghouse/Stone and Webster prototype large breeder reactor

    International Nuclear Information System (INIS)

    The optimization of the Westinghouse/Stone and Webster Prototype Large Breeder Reactor (PLBR) is described. This reactor plant, designed for ERDA and EPRI, resulted from optimization and tradeoffs on plant size, number of loops, steam cycles, system temperature, pump location, refueling concept, reactor shut-down system logic, control system logic, steam generating system, residual heat removal system, core arrangement and reactor vessel. The result is a three loop LMFBR rated at 1000 MWe gross with sodium entering the reactor vessel at 6500F (3450C) and leaving at 9500F) (5100C). The reactor vessel has a flat closure head on top, containing three rotating plugs on which are mounted the upper internals structure and the in-vessel transfer machine. The core has three radial layers of core material separating four radial blanket regions. An inclined refueling chute penetrates the reactor vessel. Plant efficiency of 37% is achieved with the use of once-through steam generators operating in the modified Sulzer mode, producing 2200 psi, 8500F (15.2 x 106 pa/4550C) steam. The residual heat removal system (RHRS) consists of three independent heat removal paths in which the intermediate sodium is cooled in air blast heat exchangers (author)

  11. Implications of nuclear physics in the development of Fast Breeder Reactors

    International Nuclear Information System (INIS)

    The purpose of this paper is to point out the involved aspects of nuclear physics in the calculation and design of the fast reactors. After a brief description of the advantages of using the fast reactors in the national economy, the national programs concerning this activity are presented. The structure and operation conditions of the liquid metal fast breeder reactor (LMFBR) are also reviewed. Then, the methods aimed to calculate the core, the burn-up, the reactor dynamics, the analysis of accidents, the shielding, as well as, the materials required in the fast reactor calculation, are shortly given. Further on, it deals with the nuclear data types connected to the fast reactor calculations, with accuracy requirements for nuclear data, as well as, with the present stage of nuclear data for fissile, fertile and structural materials. The requirements for new differential data measurements, new integral data and benchmark experiments are presented. Data adjustement methods are also summarized. Some aspects of the structural material behaviour in intense gamma radiation and neutron fields existing into a fast reactor are also presented in the last part of this paper. The concluding remarks are mentioned at the end of the paper. (author)

  12. Reactor safety equipments

    International Nuclear Information System (INIS)

    Purpose: To positively recover radioactive substances discharged in a dry well at the time of failure of a reactor. Constitution: In addition to the emergency gas treating system fitted to a reactor building, a purification system connected through a pipeline to the dry well is arranged in the reactor building. This purification system is connected through pipes fitted to the dry well to forced circulation device, heat exchanger, and purification device. The atmosphere of high pressure steam gases in the dry well is derived to the heat exchanger for cooling, and then radioactive substances which are contained in the gases are removed by filter sets charged with the HEPA filters and the HECA filters. At last, there gases are returned to dry well by circulation pump, repeat this process. (Kamimura, M.)

  13. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    International Nuclear Information System (INIS)

    The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

  14. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  15. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  16. Reactivity control system of a passively safe thorium breeder pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MWth passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm and

  17. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  18. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  19. Current status of experimental breeder reactor-II [EBR-II] shutdown planning

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor--II (EBR-II) at Argonne National Laboratory--West (ANL-W) in Idaho, was shutdown in September, 1994 as mandated by the US Department of Energy. This sodium cooled reactor had been in service since 1964, and was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the Sodium Process Facility. The sodium environment and the EBR-II configuration, combined with the radiation and contamination associated with thirty years of reactor operation, posed problems specific to liquid metal reactor deactivation. The methods being developed and implemented at EBR-II can be applied to other similar situations in the US and abroad

  20. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  1. US reactor safety study reviewed

    International Nuclear Information System (INIS)

    The policy statement issued by the United States Nuclear Regulatory Commission concerning its views of the Reactor Safety Study (WASH-1400) in the light of criticisms of the study made by a special review group headed by Dr Harold Lewis, is discussed. Dr Lewis appeared before a Congressional sub-committee to discuss the findings of the review group and his testimony is reproduced here. This is considered under the four headings; safety - what is it, WASH-1400 and RARG - what was found, risk assessment methodology -where do we go from here, and are reactors safe. (UK)

  2. Review of uncertainty estimates associated with models for assessing the impact of breeder reactor radioactivity releases

    International Nuclear Information System (INIS)

    The purpose is to summarize estimates based on currently available data of the uncertainty associated with radiological assessment models. The models being examined herein are those recommended previously for use in breeder reactor assessments. Uncertainty estimates are presented for models of atmospheric and hydrologic transport, terrestrial and aquatic food-chain bioaccumulation, and internal and external dosimetry. Both long-term and short-term release conditions are discussed. The uncertainty estimates presented in this report indicate that, for many sites, generic models and representative parameter values may be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, especially those from breeder reactors located in sites dominated by complex terrain and/or coastal meteorology, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under these circumstances to reduce this uncertainty. However, even using site-specific information, natural variability and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose or concentration in environmental media following shortterm releases

  3. Status of Fast Breeder Reactor Development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    In 1967 and 1968, the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (''DeBeNe'') agreed to develop breeder reactors in a joint program. The following research organizations have taken part in this effort: - Kernforschungszentrum Karlsruhe (KfK) - Interatom, Bergisch Gladbach Alkem, Hanau - SCK/CEN, Mol - Belgonucleaire, Brussels - ECN, Petten - TNO, Apeldoorn - Neratoom, The Hague. The first three institutions mentioned above have been associated in the Entwicklungsgemeinschaft Schneller Brüter since 1977. KfK, INTERATOM, and the French Commissariat à l'Energie Atomique entered into contracts in 1977 about close cooperation in the fast breeder field, to which the Belgian and Dutch partners acceded. The results of activities carried out by the DeBeNe partners in 1987 have been compiled in this report. The report begins with a survey of the fast reactor plants, which is followed by an R&D summary. In an additional chapter, a survey is given of international cooperation in 1987

  4. Numerical simulation of convection of argon gas in fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Turbulent convective motion of Argon gas in the annulus of fast breeder reactor. • Circumferential temperature difference (CTD) is proportional to the Rayleigh Number. • A pair of ascending and descending rolls that move azimuthally in the annulus. • Observed flow reversals. • Temperature fluctuations decrease with the height of the annulus. - Abstract: In this paper, we present the results of numerical simulations of the turbulent convection in the Argon gas present in the annulus of a fast breeder reactor. We employ RANS scheme with k–∊ model and solve the equations using an open-source software OpenFOAM. The Rayleigh numbers Ra of our simulations lie in the range of 108 to 1010. We observe a pair of rolls with a hot plume rising from one end, and a cold plume descending from the opposite end of the annulus. This feature results because the aspect ratio of the geometry is near unity. We also find that the circumferential temperature difference (CTD) is proportional to Ra

  5. Review of uncertainty estimates associated with models for assessing the impact of breeder reactor radioactivity releases

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.; Little, C.A.

    1982-08-01

    The purpose is to summarize estimates based on currently available data of the uncertainty associated with radiological assessment models. The models being examined herein are those recommended previously for use in breeder reactor assessments. Uncertainty estimates are presented for models of atmospheric and hydrologic transport, terrestrial and aquatic food-chain bioaccumulation, and internal and external dosimetry. Both long-term and short-term release conditions are discussed. The uncertainty estimates presented in this report indicate that, for many sites, generic models and representative parameter values may be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, especially those from breeder reactors located in sites dominated by complex terrain and/or coastal meteorology, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under these circumstances to reduce this uncertainty. However, even using site-specific information, natural variability and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose or concentration in environmental media following shortterm releases.

  6. Comparison of material property specifications of austenitic steels in fast breeder reactor technology

    International Nuclear Information System (INIS)

    Austenitic stainless steels are very widely used in components for European Fast Breeder Reactors. The Activity Group Nr.3 ''Materials'', within Working Group ''Codes and Standards'' of the Fast Reactor Co-Ordination Committee of the European Communities, has decided to initiate a study to compare the material property specifications of the austenitic stainless steel used in the European Fast Breeder Technology. Hence, this study would allow one to view rapidly the designation of a particular steel grade in different European countries and to compare given property values for a same grade. There were dissimilarities, differences or voids appear, it could lead to an attempt to complete and/or to uniformize the nationally given values, so that on a practical level interchangeability, availability and use ease design and construction work. A selection of the materials and of their properties has been made by the Working Group. Materials examined are Stainless Steel AISI 304, 304 L, 304 LN, 316, 316 L, 316 LN, 316''Ti stab.'', 316''Nb stab''., 321, 347

  7. Preliminary design of a Binary Breeder Reactor; Diseno preliminar de un reactor esferico de quema/cria

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  8. Research for enhancing reactor safety

    International Nuclear Information System (INIS)

    Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG)

  9. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    International Nuclear Information System (INIS)

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  10. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Budd, W.A. (ed.)

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  11. Liquid metal seal (LMS) - challenges for fast breeder test reactor (FBTR)

    International Nuclear Information System (INIS)

    In Fast Breeder Test reactor (FBTR), Liquid Metal Seal (LMS) is being used to maintain leak tightness between reactor vessel and rotating plugs. It is a eutectic mixture of 42% tin and 58% bismuth. This paper describes measurements of melting point of LMS using Differential Scanning Calorimeter (DSC), Make: Setaram; Model- 131 evo. The instrument was calibrated using Indium as standard with different heating rates, 5 °C/min, 10 °C/min, 15°C/min and 20 °C/min. The observed value of melting point was found to be in agreement with the literature value. The melting point of as received and used LMS (LMSH8, LMSH10 and LMSH12) from three locations of FBTR were studied using DSC with different heating rates as above. The results are presented and it can be clearly seen that LMS has undergone some modifications during the continuous usage in FBTR

  12. Low-order dynamic modeling of the Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    This report describes the development of a low-order, linear model of the Experimental Breeder Reactor II (EBR-II), including the primary system, intermediate heat exchanger, and steam generator subsystems. The linear model is developed to represent full-power steady state dynamics for low-level perturbations. Transient simulations are performed using model building and simulation capabilities of the computer software Matrixx. The inherently safe characteristics of the EBR-II are verified through the simulation studies. The results presented in this report also indicate an agreement between the linear model and the actual dynamics of the plant for several transients. Such models play a major role in the learning and in the improvement of nuclear reactor dynamics for control and signal validation studies. This research and development is sponsored by the Advanced Controls Program in the Instrumentation and Controls Division of the Oak Ridge National Laboratory. 17 refs., 67 figs., 15 tabs

  13. Single assembly preliminary analysis for horizontal seismic analysis on fast breeder reactor core

    International Nuclear Information System (INIS)

    Seismic analysis is one of important parts of fast breeder reactor (FBR) core design. It is necessary for structural integrity assessment and analysis of variation of reactivity under the earthquake. Moreover some important data for qualification of the scram capability of the control rods during the earthquake. Moreover some important data for qualification of the scram capability of the control rods during the earthquake could be provided. In the paper, FINAS, one finite element code developed by Japanese engineers, was used. The calculation model and method were studied on single assembly in China Experimental Fast Reactor (CEFR), as an example. Some preliminary analyses were carried out, which prepare for the seismic analysis on multiple assemblies in FBR core. (authors)

  14. Experimental Breeder Reactor II (EBR-II), instrumentation for core surveillance

    International Nuclear Information System (INIS)

    The paper describes the Experimental Breeder Reactor-2 (EBR-2), thermal-hydraulic testing on the facility, and features of EBR-2 subassembly design. It is reported that during 25 years of EBR-2 operation, several of original, non-replaceable flow-sensors and thermocouples have failed in the primary system, and that this has led to the development of new sensors. The conclusion is made that from test series of measurements of temperature and flow in subassemblies, EBR-2 calculations showed that the core could withstand a loss-of-flow without scram accident and a loss-of-heat sink without scram accident from full reactor power without core damage. 11 refs, 9 figs

  15. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  16. Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Highlights: → We model design optimization of a vital reactor component using Genetic Algorithm. → Real-parameter Genetic Algorithm is used for steam condenser optimization study. → Comparison analysis done with various Genetic Algorithm related mechanisms. → The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

  17. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    International Nuclear Information System (INIS)

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs

  18. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    International Nuclear Information System (INIS)

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences

  19. Equipment cell liners for liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Concepts and practices used in the design of equipment cell liners for liquid-metal-cooled fast breeder reactor (LMFBR) systems were surveyed to ascertain the manner by which the functional requirements were satisfied, the severity of sodium spills the liners were designed to accommodate, and the problems encountered in design and construction. The survey was limited to ''loop-type'' LMFBRs, with primary interest on recently constructed plants. Steel-lined concrete structures are discussed; cell-liner designs used in several LMFBR plants are described with particular emphasis on the Southwest Experimental Fast Oxide Reactor (SEFOR), which uses a fixed liner, and the Fast Flux Test Facility (FFTF), which uses a free-floating liner; and research and development work believed necessary to permit a rational and thorough assessment of cell-liner design concept is identified

  20. Significance of the SNR 300 fast breeder reactor in terms of research policy

    International Nuclear Information System (INIS)

    The publication consists of the following documents: (1) Compact version of the expert opinion on the benefit to be gained for the research policy of the FRG from the FBR prototype reactor station SNR 300. (2) Speech of the Federal Minister for Research and Technology, Dr. Heinz Riesenhuber, in the German Bundestag (September 22, 1988). (3) Survey of fast breeder reactor development and of the SNR 300. (4) Statement concerning the proposal to use the Kalkar nuclear power station (SNR 300) as a 'waste management facility' for plutonium and other actinides. (5) Reply of the Federal Government to an interpellation filed by the deputies Mr Wetzel, Mr Stratmann, Mrs Tauber, Dr. Daniels (Regensburg), and the parliamentary party of the Greens. (orig./UA)

  1. Fast current pulse amplifier for neutron flux monitoring system of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The neutron flux monitoring system (NFMS) for Prototype Fast Breeder Reactor (PFBR) measures the neutron power and the reactivity changes in the core in all the states such as shut down, fuel handling, reactor startup, intermediate and power ranges using high temperature cylindrical fission chambers, four section fission counter and high temperature boron coated counter. Fast Current Pulse Amplifier has been developed to use in NFMS of PFBR that amplifies single/four numbers of input current pulses independently, discriminates and electronically wire - OR them to give differential pulse output along with the Campbell output. The paper describes the design, development of integrated single/Quad channel fast current pulse amplifier based on in-house developed ASIC, Hybrid IC, in built test features, LV and HV supplies. (author)

  2. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  3. Development of remotely controlled in-service inspection equipment for fast breeder reactor vessels

    International Nuclear Information System (INIS)

    For the operation of fast breeder reactors, one of the most important aspects is the need to control the functioning of the components. It is a characteristic of FBRs that the reactor vessel and in particular the guard vessel operate under very severe conditions. Therefore, an improved remotely controlled inspection system would be needed. On the basis of its experience with light water reactors, Mitsubishi Heavy Industries (MHI) has developed versatile systems for in-service inspection (ISI) of the reactor vessel and its ancillary components. The paper describes what could become the most important part of the ISI system, namely a special mobile vehicle or robot, called the MOLE (Mitsubishi Original LocomobilE). This vehicle can run freely over the annulus sections of the reactor vessel and its guard vessel and can carry out various tests. The results so far have been satisfactory and have encouraged MHI to make further tests in order to confirm that the access to the vessel is sufficient for performing the necessary inspections. (author)

  4. Meeting evolving reactor safety requirements

    International Nuclear Information System (INIS)

    This paper gives an overview of the work being conducted by the Electric Power Research Institute (EPRI) in the reactor safety field. Topics covered include nuclear seismic research, nuclear reload management, systems reliability analysis, severe accident management, safety margins and testing (i.e., water hammer), source term investigations and control and diagnostics. The general philosophy behind the work is to be of help to the utilities which support EPRI in such a way that safety is enhanced, firstly, in protecting the public and, secondly, in protecting the very large investment represented by a nuclear power plant

  5. Conceptual Design Studies of a Passively Safe Thorium Breeder Pebble Bed Reactor

    OpenAIRE

    Wols, F.J.

    2015-01-01

    Nuclear power plants are expected to play an important role in the worldwide electricity production in the coming decades, since they provide an economically attractive, reliable and low-carbon source of electricity with plenty of resources available for at least the coming hundreds of years. However, the design of nuclear reactors can be improved significantly in terms of safety, by designing reactors with fully passive safety systems, and sustainability, by making more efficient use of natu...

  6. Determination of research reactor safety parameters by reactor calculations

    International Nuclear Information System (INIS)

    Main research reactor safety parameters such as power density peaking factors, shutdown margin and temperature reactivity coefficients are treated. Reactor physics explanation of the parameters is given together with their application in safety evaluation performed as part of research reactor operation. Reactor calculations are presented as a method for their determination assuming use of widely available computer codes. (author)

  7. Numerical analysis of grid plate melting after a severe accident in a Fast-Breeder Reactor (FBR)

    Indian Academy of Sciences (India)

    A Jasmin Sudha; K Velusamy

    2013-12-01

    Fast breeder reactors (FBRs) are provided with redundant and diverse plant protection systems with a very low failure probability (<10-6/reactor year), making core disruptive accident (CDA), a beyond design basis event (BDBE). Nevertheless, safety analysis is carried out even for such events with a view to mitigate their consequences by providing engineered safeguards like the in-vessel core catcher. During a CDA, a significant fraction of the hot molten fuel moves downwards and gets relocated to the lower plate of grid plate. The ability of this plate to resist or delay relocation of core melt further has been investigated by developing appropriate mathematical models and translating them into a computer code HEATRAN-1. The core melt is a time dependent volumetric heat source because of the radioactive decay of the fission products which it contains. The code solves the nonlinear heat conduction equation including phase change. The analysis reveals that if the bottom of grid plate is considered to be adiabatic, melt-through of grid plate (i.e., melting of the entire thickness of the plate) occurs between 800 s and 1000 s depending upon the initial conditions. Knowledge of this time estimate is essential for defining the initial thermal load on the core catcher plate. If heat transfer from the bottom of grid plate to the underlying sodium is taken into account, then melt-through does not take place, but the temperature of grid plate is high enough to cause creep failure.

  8. Design of fuel fabrication plant of Fast Reactor Fuel Cycle Facility for reload requirement of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India's economic growth is on a fast growth track. The energy demand is expected to grow rapidly in the coming decades. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear energy is best suited to meet this demand in a sustainable manner without causing undue environmental impact. Fast reactors are expected to be major contributors in sufficing this demand to a great extent. As an effort to achieve the objective, a Prototype Fast Breeder Reactor is being constructed at Kalpakkam. This paper also highlights the design features of FFP, unit operations, scheme of automation, branched layout of glove box train, shielding arrangement on glove boxes, accident consequence analysis etc.

  9. The safety of light water reactors

    International Nuclear Information System (INIS)

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  10. Status and prospects of thermal breeders and their effect on fuel utilization

    International Nuclear Information System (INIS)

    The report evaluates the extent to which thermal breeders and near-breeders might complement fast breeders or serve as an alternative in solving the long-term nuclear fuel supply problem. It considers in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability and fuel cycle versatility in order to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  11. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    International Nuclear Information System (INIS)

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C β-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ∼3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation

  12. Development of advanced blanket materials for solid breeder blanket of fusion reactor

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  13. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)

  14. Fusion breeder

    International Nuclear Information System (INIS)

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

  15. Safety of thermal water reactors

    International Nuclear Information System (INIS)

    This book reports on the latest European research into the safety of thermal water reactors, based on the presentation and evaluation of results obtained from research projects undertaken in different national laboratories of the European Community. Information is included under the following areas of research: 1.) The loss of coolant accident (LOCA) and the functioning and performance of the emergency core cooling system; 2.) The protection of nuclear power plants against external gas cloud explosions; and 3.) The release and distribution of radioactive fission products in the atmosphere following a reactor accident

  16. Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM

  17. Global depletion analysis of Korean helium cooled solid breeder TBM model for demo fusion reactor

    International Nuclear Information System (INIS)

    The Korean HCSB (helium cooled solid breeder) TBM (test blanket module) is proposed with its specific compositions of lithium ceramic, beryllium and graphite in pebble form. In the Korean HCSB TBM, the amount of beryllium is reduced and the reduction is replaced by graphite for a neutron reflector, while tritium breeding ratio (TBR) remains almost unchanged with relatively low Li6 enrichment of ∼40%. However, the previous Korean HCSB was designed based on the LOCAL assumption, in which the surroundings are assumed by the reflective boundary condition. In this research, we establish a simple GLOBAL neutronics model based on demo fusion reactor and perform neutronics analyses including depletion (transmutation) calculation during 100 EFPDs (effective full power days) using the modified MONTEBURNS code.

  18. Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

    International Nuclear Information System (INIS)

    The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D2 gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment. (authors)

  19. Development of an ISI Robot for the Fast Breeder Reactor MONJU Primary Heat Transfer System Piping

    International Nuclear Information System (INIS)

    This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire type ultrasonic sensor for volumetric tests at high temperature (atmosphere 55 degree C, Piping Surface 80 degree C) and radiation exposure condition (dose rate 10 mGy/h, piping surface dose rate 15 mGy/h). It was developed an inspection robot using a new tire type for the ultrasonic testing sensor and a new control method. A signal to noise ratio S/N over 2 was obtained during the functional test for a calibration defect with depth 50%t (from the tube wall thickness). (author)

  20. Ultrasonic inspection of liquid-metal fast breeder reactor steam generator duplex tubing

    International Nuclear Information System (INIS)

    Two ultrasonic inspections of the Experimental Breeder Reactor II steam generator duplex tubing have been completed. Inspections performed on one evaporator in 1976 provided baseline data, and a subsequent inspection in 1978 revealed no change in tube condition. With the completion of the 1978 inspection, all available tubes in one evaporator have been inspected. The steam generator contains duplex tubes fabricated from 2 1/4 Cr-1 Mo ferritic steel. Access to the bore (water) side of the tubes was gained through the steam outlet piping. The inspection included a complete volumertic (100% of the tube material) examination, measurement of wall thickness, and evaluation of the condition of the braze bonding the two walls of the tube together. The test equipment was routinely calibrated against a standard containing artificial flaws. Artificial flaws as small as 1.6 mm long x 0.25 mm deep were readily detected

  1. C-scope under-sodium viewer for sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    A C-scope under-sodium viewer has been developed for monitoring the interior of sodium-cooled fast breeder reactors. Consisting of a transducer that emits and receives ultrasonic waves under liquid sodium, a mechanism that drives the transducer under liquid sodium and an image displaying section, it inspects the fuel assembly through its image in optically opaque high-temperature (3000C) liquid sodium. The results of its evaluation test are: (1) The transducer could continue satisfactory operation under 3500C (at the highest) sodium for more than a month. (2) The driving mechanism, though it was the first of the kind appearing in Japan, has been proved that it could continue operation for a week under 3000C sodium. (3) The image displaying section, in spite of the low speed of the transducer (below 20 rpm), could display stable and clear images. (4) The image in 3000C was as clear as that in room-temperature water. (auth.)

  2. Clinch River Breeder Reactor: an assessment of need for power and regulatory issues

    Energy Technology Data Exchange (ETDEWEB)

    Hamblin, D M; Tepel, R C; Bjornstad, D J; Hill, L J; Cantor, R A; Carroll, P J; Cohn, S M; Hadder, G R; Holcomb, B D; Johnson, K E

    1983-09-01

    The purpose of this report is to present the results of a research effort designed to assist the US Department of Energy in: (1) reviewing the need for power from the Clinch River Breeder Reactor (CRBR) in the Southeastern Electric Reliability Council (SERC) region, not including Florida, and (2) isolating specific regulatory and institutional issues and physical transmission capacities that may constrain the market for CRBR power. A review of existing electric power wheeling arrangements in the Southeast and specific federal and state regulatory obstacles that may affect power sales from the CRBR was undertaken. This review was a contributing factor to a decision to target the service territory to SERC-less Florida.

  3. Gamma-ray spectra of fast-breeder spent nuclear fuel from the BN-350 reactor

    International Nuclear Information System (INIS)

    Gamma-ray measurements of spent nuclear fuel (SNF) from a fast breeder reactor have been obtained with a High-Purity Germanium (HPGe) Detector. The HPGe measurements were performed inside a hot cell using an adjustable collimator to restrict the viewing angle of the HPGe to a small region of the SNF assembly. In addition Ion Chamber (IC) measurements were performed underwater using a lead shielded IC 15-cm in active length. We are going to present HPGe measurement results of the distribution of fission product and activation products along the assembly. We will also compare the gamma-ray profiles of the HPGe and IC measurements to those of the neutron profiles measured with a 3 He tube based neutron counter

  4. Development of high temperature fission counter-chamber(FC)S for a top entry loop type fast breeder reactor

    International Nuclear Information System (INIS)

    Prototype high temperature fission counter-chambers have been made as neutron detectors for installation in the reactor vessel of the 600MWe-class top entry loop type fast breeder reactor. Using these prototypes as samples, a high-temperature endurance test has been conducted. The validity of the prototypes has been established by the test results, which show that the prototypes nearly satisfy the design performance. (author)

  5. Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

    International Nuclear Information System (INIS)

    Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work

  6. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  7. Development of a nuclear information system for the MONJU Fast Breeder Reactor

    International Nuclear Information System (INIS)

    At the MONJU Fast Breeder Reactor information is collected on a continuous basis. This information consists of measured data, design data, simulations data, maintenance data etc. which may be in any kind of electronic form, i.e. text documents, code input files, reports or even scanned documents. The amount and nature of these data has introduced the need for a software system, which will provide an efficient infrastructure for the maintenance of and operations on the data. Thus a Nuclear Information System for the MONJU Fast Breeder Reactor is under development. The system consists of remote databases hosting the information along with clients for handling them, remote clients providing the users with an interface and a local server for handling the client requests and the communication between the database and user clients. The system is composed of independent server, database and user modules, which communicates using the RMI-IIOP (Remote Method Invocation - Internet InterORB Protocol) technology. The RMI-IIOP is a CORBA (Common Object Request Broker Architecture) compliant subset of the RMI thereby facilitating the possibility of implementing the database and user modules in any kind of programming language and on any kind of operating system by providing a standard, platform independent communications interface. The user interface consists of dynamic HTML web pages which instantiates servlets in the user module when the user submits queries. The database module consists of controllers for handling the communication with the user module and database drivers for handling the connections with the databases. In this paper the overall system design and schemes for data flow and remote method invocations are presented and the requirements imposed on the system are discussed. (author)

  8. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    International Nuclear Information System (INIS)

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected

  9. Design and fabrication of steam generators (superheaters) for the prototype fast breeder reactor 'MONJU'

    International Nuclear Information System (INIS)

    In liquid metal-cooled fast breeder reactors, steam generators are one of the important equipments, and emphasis has been placed on their development in various countries in the world. Also in Japan, centering around the Power Reactor and Nuclear Fuel Development Corp., the research and development in the wide range from the fundamentals on heat transfer and flow, materials and strength for steam generators to the manufacture, operation and various tests of large mock-ups including a 50 MW steam generator have been carried out. Further, as for the manufacture and inspection, the improvement of the method of welding tubes and tube plates, the adoption of a fine focus X-ray inspection apparatus and others were carried out. Moreover, as the maintenance technique, the ultrasonic flaw detection probes for the heating tubes were developed. The steam generators (superheaters) for the FBR 'Monju' power station are the heat exchangers of helical coil tube-shell type using SUS 321 steel as the heating tube material. Based on the results of these research and development, the design and manufacture of these superheaters and their installation in the reactor auxiliary building of the FBR 'Monju' power station were completed. The outline of the design, the research and development and the manufacture of the steam generators (superheaters) are reported. (K.I.)

  10. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  11. Fabrication and quality control of MOX fuel for Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Full text: Uranium-Plutonium mixed oxide (MOX) fuel for both thermal and fast reactors have been fabricated by Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, India. MOX fuel bundles fabricated by AFFF have been loaded in Boiling Water Reactors (BWRs) and Pressurised Heavy Water Reactors (PHWRs) and have been discharged after successful irradiation. An experimental fuel subassembly containing 37 MOX pins is being irradiated in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai and has seen a burn up of more than 80000 MWD/T. MOX fuel pins containing 44% Pu02 have been recently loaded as a part of the hybrid core of FBTR. AFFF has now taken up the manufacture of MOX fuel pins for the Prototype Fast Breeder Reactor (BHAVINI) coming up at Kalpakkam. The core consists of 181 sub assemblies containing 217 MOX fuel pins each. It is required to fabricate nearly 40,000 MOX fuel pins (3 meter long) for the first core. The Prototype Fast Breeder Reactor is designed with two different fissile enrichment zones to be loaded with MOX subassemblies with a nominal composition of 21% and 28% of PuO2. The fuel pellets of required composition are made using conventional powder metallurgy processes. The pellets are annular with an inner hole of 1.8mm diameter and outside diameter of 5.5mm. AFFF has developed the technology of making annular MOX fuel pellets for PFBR and optimized conditions of fabrication. Multistation rotary presses have been used for compaction of the pellets. The fuel pin consists of a MOX stack of 1000mm and axial blanket of deeply depleted uranium dioxide of length 300mm on either side. New techniques have been used at different stages of fabrication of the fuel pins namely pelletisation, welding and wire wrapping. Studies have been made to use laser welding technique for welding of endplugs. Automation has been introduced in a number of process steps in the fabrication line. A detailed quality control plan is prepared

  12. Fabrication and quality control of MOX fuel for Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Uranium-Plutonium mixed oxide (MOX) fuel for both thermal and fast reactors have been fabricated by Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, India. MOX fuel bundles fabricated by AFFF have been loaded in Boiling Water Reactors (BWRs) and Pressurised Heavy Water Reactors (PHWRs) and have been discharged after successful irradiation. An experimental fuel subassemby containing 37 MOX pins is being irradiated in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai and has seen a burn up of more than 92000 MWd/t. MOX fuel pins containing 44% PuO2 have been recently loaded as a part of the hybrid core of FBTR. AFFF has now taken up the manufacture of MOX fuel pins for the Prototype Fast Breeder Reactor (PFBR) coming up at Kalpakkam . The core consists of 181 sub assemblies containing 217 MOX fuel pins each. Prototype Fast Breeder Reactor is designed with two different fissile enrichment zones to be loaded with MOX subassemblies with a nominal composition of 21% and 28% of PuO2.The fuel pellets of required composition are made using conventional powder metallurgy processes. The pellets are annular with an inner hole of 1.8 mm diameter and outside diameter of 5.5 mm. AFFF has developed the technology of making annular MOX fuel pellets for PFBR and optimized conditions of fabrication. Multistaion rotary presses have been used for compaction of the pellets. The fuel pin consists of a MOX stack of 1000 mm and axial blanket of deeply depleted uranium dioxide of length 300 mm on either side. New techniques have been used at different stages of fabrication of the fuel pins namely pelletisation, welding and wire wrapping. Studies have been made to use laser welding technique for welding of endplugs. Automation has been introduced in a number of process steps in the fabrication line. A detailed quality control plan is prepared based on the specifications and advanced process and quality control procedures have been incorporated to

  13. Safety system of reactor container

    International Nuclear Information System (INIS)

    The safety system of the present invention can shut down a BWR type reactor safely without operator's aid even upon occurrence of an abnormal state. Namely, a pressure/temperature measuring and controlling device is disposed to a dry well in the safety system of a reactor container incorporating a pressure vessel, a space between a dry wall and a wet well and a pressure suppression chamber. Operation signals sent from pipelines of an emergency reactor core cooling system delivered from the pressure vessel are inputted to the pressure/temperature measuring and controlling device. Output signals of the pressure/temperature measuring and controlling device are inputted to a spray device. With such procedures, when actuation of dry well spray is required upon loss of coolants accident, necessity for the actuation of the spray can be judged based on the pressure, temperature in the dry well, reactor water level and a state of operation and duration of abnormal state of other ECCS system using the pressure/temperature measuring and controlling device disposed in the dry well. If actuation of spray is required, the dry wall spray is automatically actuated to reduce pressure and temperature in the container. (I.S.)

  14. Reactor safety research in Sweden

    International Nuclear Information System (INIS)

    Objectives, means and results of Swedish light water reactor safety research during the 1970s are reviewed. The expenditure is about 40 Million Swkr per year excluding industry. Large efforts have been devoted to experimental studies of loss of coolant accidents. Large scale containment response tests for simulated pipe breaks were carried out at the Marviken facility. At Studsvik a method for testing fuel during fast power changes has been developed. Stress corrosion, crack growth and the effect of irradiation on the strength ductility of Zircaloy tube was studied. A method for determining the fracture toughness of pressure vessel steel was developed and it was shown that the fracture toughness was lower than earlier assumed. The release of fission products to reactor water was studied and so was the release, transport and removal of airborne radioactive matter for Swedish BWRs and PWRs. Test methods for iodine filter systems were developed. A system for continuous monitoring of radioactive noble gas stack release was developed for the Ringhals plant. Attention was drawn to the importance of the human factor for reactor safety. Probabilistic methods for risk analysis were applied to the Barsebaeck 2 and Forsmark 3 boiling water reactors. Procedures and working conditions for operator personnel were investigated. (GBn)

  15. Nuclear reactors. To breed or not to breed. A Pugwash debate on fast breeder reactors held at the Royal Society, London, on 28 September 1976 under the chairmanship of Sir Alec Merrison

    International Nuclear Information System (INIS)

    The debate which is reported was timed to coincide with the publication of the Report of the (UK) Royal Commission on Environmental Pollution: 'Nuclear Power and Environment'. The volume comprises an introductory section, a report of an address by the Chairman of the Royal Commission and invited papers on fast breeder reactors in relation to energy requirements, on the safety of a commercial fast reactor, on processing and reprocessing of fuel, on radioactive waste management, and on diversion of plutonium and proliferation of nuclear weapons. An edited version of the discussion is presented under the following heads: comments on the report of the Royal Commission; projections of future energy requirement; thermal pollution; safety and insurance of reactors; reprocessing of fuel; storage and disposal of wastes; energy from fusion; utilization of coal; and proliferation of weapons and diversion of plutonium. The six invited papers are considered to be within INIS scope and separate abstracts have been prepared. (U.K.)

  16. Advanced reactor concepts and safety

    International Nuclear Information System (INIS)

    The need for some consistency in the terms used to describe the evolution of methods for ensuring the safety of nuclear reactors has been identified by the IAEA. This is timely since there appears to be a danger that the precision of many valuable words is being diluted and that a new jargon may appear that will confuse rather than aid the communication of important but possibly diverse philosophies and concepts. Among the difficulties faced by the nuclear industry is promoting and gaining a widespread understanding of the risks actually posed by nuclear reactors. In view of the importance of communication to both the public and to the technical community generally, the starting point for the definition of terms must be with dictionary meanings and common technical usage. The nuclear engineering community should use such words in conformance with the whole technical world. This paper addresses many of the issues suggested in the invitation to meet and also poses some additional issues for consideration. Some examples are the role of the operator in either enhancing or degrading safety and how the meaning or interpretation of the word 'safety' can be expected to change during the next few decades. It is advantageous to use criteria against which technologies and ongoing operating performance can be judged provided that the criteria are generic and not specific to particular reactor concepts. Some thoughts are offered on the need to frame the criteria carefully so that innovative solutions and concepts are fostered, not stifled

  17. Evaluation of the Community's nuclear reactor safety research programme

    International Nuclear Information System (INIS)

    This report describes an evaluation of the 1980-85 CEC reactor safety programme prepared, at the invitation of the Commission, by a panel of six independent experts by means of examining the relevant document and by holding hearings with the responsible CEC staff. It contains the recommendations made by the panel on the following topics: the need for the JRC to continue to make its competence in the reactor safety field available to the Community; the importance of continuity in the JRC and shared-cost action programmes; the difficulty of developing reactor safety research programmes which satisfy the needs of users with diverse needs; the monitoring of the utilization of the research results; the maintenance of the JRC computer codes used by the Member States; the spin-off from research results being made available to other industrial sectors; the continued contact between the JRC researchers and the national experts; the coordination of LWR safety research with that of the Member States; and, the JRC work on fast breeders to be planned with regard to the R and D programmes of the Fast Reactor European Consortium

  18. Safety issues at the defense production reactors

    International Nuclear Information System (INIS)

    The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors - the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. This report examines the safety objective established by the Department of Energy for the production reactors and the process the Department of its contractors use to implement the objective; focuses on a variety of uncertainties concerning the production reactors, particularly those related to potential vulnerabilities to severe accidents; and identifies ways in which the DOE approach to management of the safety of the production reactors can be improved

  19. Development of fluorocarbon rubber for backup seals of sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: → Negligible chemical degradation of seal compound during ageing (in unstrained state) in air at 140/170/200 oC for 32 weeks. → Cross-link exchange, Joule-Gough effect and ionic interaction during ageing in unstrained state. → Enhanced physical/chemical degradation of compound during ageing under strain. → Capability of compound to withstand heat, radiation, air and mechanical load in reactor for 10 years. → Negligible chemical dose rate effect and gas evolution from compound during seal operation. -- Abstract: The development of a fluorohydrocarbon rubber compound for static backup seals of 500 MWe, Prototype Fast Breeder Reactor (PFBR) is depicted. Variations of a previously developed Viton A-401C based formulation were subjected to processability tests, accelerated heat ageing in air, mechanical characterization and production trials. Finite element analysis and literature data extrapolation were combined with long term ageing to ascertain the life (minimum 10 years) of chosen formulation in reactor under synergistic influences of 110 oC, 23 mGy/h (γ dose rate) and air considering postulated accidental conditions. Validation of test seals and quality assessment indicate that composition and properties of the validated laboratory compound has been translated effectively to the reactor seals, installed recently in PFBR. The tensile and hardness specimens indicated negligible degradation and exceptional thermo-oxidative stability of the seal compound during ageing (32 weeks at 140/170/200 oC) even though interesting manifestations of cross-link exchange and ionic interactions were observed. Compression set results, showing definite trends of change under ageing and stain, were used in Arrhenius and Williams Landel Ferry equations for realistic life prediction. The development provides a foundation to simplify and standardize the design, development and operation of major elastomeric sealing applications of Indian nuclear reactors based on a

  20. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  1. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. Pt. II. R and D necessities and development across the world. A review

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions Coimbatore, Tamilnadu (India)

    2013-12-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative study of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. R and D necessities for inflatable seals and their development across the world are given closer look in Part II of the review in continuation of Part I. (orig.)

  2. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. Pt. II. R and D necessities and development across the world. A review

    International Nuclear Information System (INIS)

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative study of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton registered GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. R and D necessities for inflatable seals and their development across the world are given closer look in Part II of the review in continuation of Part I. (orig.)

  3. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and The Netherlands

    International Nuclear Information System (INIS)

    The results of activities carried out by the DeBeNe partners in 1989 have been compiled in this report. The report begins with a survey of fast reactor plants, which is followed by an R and D summary. In an additional chapter, a survey is presented of international cooperation in 1989. Effective January 1990, KfK activities in the area of fast reactors have been incorporated in the Nuclear Safety Research Project. (author)

  4. Effect of geometric factors on performance of a sodium to air heat exchanger in a fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • A heat exchanger analysis (HE) before scale up reduces excess heat transfer area. • Representative Elementary Volume analysis of a HE speeds up the solution. • The error in air temperature rise prediction by numerical across HE is within 5%. • When both pitches are reduced, the maximum increase in heat flux is experienced. • The experience has resulted in better design of next level heat exchangers. - Abstract: Prototype fast breeder reactor (PFBR) has a safety grade decay heat removal system whose performance depends on the effective functioning of natural convection heat exchangers called sodium to air heat exchangers. The development of Representative Elementary Volume (REV) model for the sodium to air heat exchanger is necessary to envisage its design and to study the effect of various factors for continuous improvement in design. With a Representative Elementary Volume, the hydrodynamic and heat transfer characteristics of the heat exchanger was studied and the results agree well with experimental data. The effect of longitudinal pitch and transverse pitch on the heat exchanger performance has been studied and an improvement of 22% in heat transfer is predicted

  5. The safety of future reactors

    International Nuclear Information System (INIS)

    To sum up, I would like to underline once again the importance of experience feedback. This issue can only be properly handled by reversing the thought process which lay behind the construction of the current NPP's. The design was the springboard for building the reactors and then operating them. Throughout construction and at times during operation, many difficulties arose, which were overcome by modifications. The need today is to go back down the line in the opposite direction : to use operational and constructional experience to restructure the design. Furthermore, the design of future reactors appears to me as a process which must be founded upon two guiding principles : defense in depth and a PSA-type probabilistic approach. They seem to me ideally fitted to underpin such a process, especially in the case of an evolutionary-type reactor project. Such a strategy requires the cooperation of many participants supported by a high level of safety culture, as defined in the report published by the IAEA in 1991 : a permanent questioning attitude, a prudent approach and efficient communication between all of the individuals and organizations involved. Failure to make such an effort might well compromise the safety goals mentioned earlier in this paper. (author) any other organization. (author)

  6. Fabrication of MOX Fuel elements for irradiation in Fast Breeder Test Reactor (FBTR)

    International Nuclear Information System (INIS)

    Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur is fabricating Uranium - Plutonium Mixed Oxide Fuel (MOX) for different types of reactors. Recently MOX fuel pins for an experimental fuel subassembly of 37 pins has been fabricated for irradiation in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai. MOX fuel pins containing 44% PuO2 have also been also made for the hybrid core of FBTR. The experimental sub-assembly for irradiation testing in FBTR consisted of 37 short length Prototype Fast Breeder Reactor (PFBR) MOX fuel elements. The composition of the fuel was (0.71 U - 0.29 Pu) O2 with U233 O2 content of 53.5% of total UO2. Uranium enriched with U233 was used to simulate the heat flux of PFBR in FBTR neutron spectrum. MOX fuel pellets were made by powder metallurgy process consisting of pre-compaction, granulation, final compaction and sintering at high temperature. Initially U3233 O8 / U233 O3 powder was subjected to heat treatment. The pellets were sintered at reducing atmosphere at 1650oC for 4 hours to obtain acceptable quality pellets. Over sized pellets were centrelessly ground.without using a liquid coolant. During the fabrication of pins for experimental subassembly, technology was developed and conditions were optimized for making annular pellets, TIG welding of D9 tubes with SS 316 end plugs and wire wrapping. Quality control procedures and process control procedures at different stages of fabrication were developed. The hybrid core of FBTR consists of Mixed Carbide (MC) sub-assemblies containing (0.70 Pu - 0.30 U) C pellets and MOX fuel sub-assemblies containing (0.44 Pu - 0.56 U) O2. Studies were made to fabricate fuel containing higher percentage of Plutonium and the conditions were established. This paper describes the development of flowsheet for making annular MOX fuel pellets containing plutonium and U233, the technology for welding of D-9 clad tubes, wire wrapping and inspection. The paper also

  7. Status of fast breeder reactors and associated fuel cycle in India

    International Nuclear Information System (INIS)

    Full text: India is the largest democracy with the current population of about 1.05 billion and is on a road to rapid growth in economy. An impressive average domestic product (GDP) growth rate of about 8 % per year has been achieved in 2006-07 and it is targeted to touch 10 % per year for the next 10 years. Towards realizing this targeted growth, development activities are planned based on well-conceived road map and clear vision. Like elsewhere, the energy and electricity growth in India are also closely linked to growth in economy. Indices of socio-economic development like literacy, longevity, GDP and human development are directly dependent upon the per capita energy consumption of a country. India is aiming to reach at least per capita energy consumption equal to the present world average (2200 kWh/a) by 2030 from the current value of (660 kWh/a). The current installed capacity of ∼138 GW(e) needs to be increased to about 600 GWe by 2030 assuming the population of about 1.4 billion. Energy strategists in the country have realized the importance of judicious mix of energy resources to meet this challenge. A large share of nuclear energy is an inevitable choice in this judicious energy mix from resources, sustainability and environment considerations. The nuclear is expected to contribute about 63 GWe by 2030, which will be steadily increased to 275 GWe by 2052, against the total projected capacity of 1344 GWe. The three stage visionary programme of India envisages Water Reactors (first stage), Fast Breeders with high breeding (second stage) and Thorium based Reactors as third stage. Closed fuel cycle in all stages is an essential ingredient. The success of each stage depends upon expeditious maturity of the earlier stage as India has limited indigenous resources of uranium, but vast resources of thorium. India ranks high in nuclear technology scale with strong R and D, high quality human resources, sound infrastructure, unwavering Government support and

  8. Present state of new technologies of nuclear power generation, and technological development of fast-breeder reactor and next-generation light water reactor

    International Nuclear Information System (INIS)

    This paper introduces the present state of development of FBR in Japan and international cooperation, the development of HP-ABWR and HP-APWR as the next-generation light water reactors, and SMR development in the United States. As for FBR, the following situations are described: (1) history of development in Japan in the past, (2) history of change due to the readjustment of development plan caused by the accident of Fukushima Daiichi Nuclear Power Station, in which shift to FaCT phase 2 was suspended, and the approach to the establishment of safety standards for sodium-cooled FBR and its international standardization was adopted, and (3) future challenges. As for the Japan - France fast-breeder reactor development cooperation, the conclusion of the Japan - France inter-government agency agreement, and Japan's cooperation plan and system are described. Next, as for HP-ABWR and HP-APWR, the development goal and concept of each plant, and the element technologies required for the success are described. On the other hand, the small reactor development in the United States started with the aim of the securement of domestic technology base, contribution to reduction in carbon dioxide emissions, and its export to new entry countries for nuclear energy. This project aimed the practical use of SMR, and started 'financial support program for small reactors' to allocate about 452 million dollars to maximum two units of SMRs in the next five years. This project is outlined. (A.O.)

  9. Advanced automation concepts applied to Experimental Breeder Reactor-II startup

    International Nuclear Information System (INIS)

    The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs

  10. Real Time Computer for Plugging Indicator Control of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Liquid sodium is used as coolant to transfer the heat produced in the reactor core to steam water circuit. Impurities present in the sodium are removed using purification circuit. Plugging indicator is a device used to measure the purity of the sodium. Versa Module Europa bus based Real Time Computer (RTC) system is used for plugging indicator control. Hot standby architecture consisting of dual redundant RTC system with switch over logic system is the configuration adopted to achieve fault tolerance. Plugging indicator can be controlled in two modes namely continuous and discontinuous mode. Software based Proportional-Integral-Derivative (PID) algorithms are developed for plugging indicator control wherein the set point changes dynamically for every scan interval of the RTC system. Set points and PID constants are kept as configurable in runtime in order to control the process in very efficient manner, which calls for reliable communication between RTC system and control station, hence TCP/IP protocol is adopted. Performance of the RTC system for plugging indicator control was thoroughly studied in the laboratory by simulating the inputs and monitored the control outputs. The control outputs were also monitored for different PID constants. Continuous and discontinuous mode plots were generated. (authors)

  11. Considerations of the effects of azimuthal fuel motion in a fast breeder reactor accident

    International Nuclear Information System (INIS)

    A sizeable reactivity feedback can result from material movement in a large liquid metal fast breeder reactor (LMFBR). Previous investigations considered mainly fuel slumping under gravity and outward radial motion. Very little work has been done on azimuthal motion. Furthermore, studies of the effects of material motion either use point kinetics or very expensive space-time differencing solutions. This work was undertaken to develop an intermediate approach between point kinetics and the full space-time finite difference solutions. The approach was then applied to sample problems with azimuthal and radial material motion under possible accident conditions. Specifically, the objectives of the work were to: (a) develop a technique for treating space-time neutron kinetics during postulated accident conditions which include material motion; (b) apply the technique to sample problems. The technique was developed based on the use of the finite element method (FEM) for the spatial differencing of the multigroup, time dependent diffusion equation. The FEM was chosen for three reasons. First, the FEM gives good accuracy with relatively fewer unknowns than the finite difference method. Second, the FEM is very flexible in setting up a mesh for the geometry of concern. Last, the FEM could handle the spatial differencing of a mesh which became distorted as the material in the reactor moved. This material motion was handled by specifying the FEM mesh nodes as a function of time and periodically updating the spatial matrices. Finally, the method used to solve the time dependence was Gear's variable order predictor corrector scheme

  12. Design and fabrication of sodium test facility for fast breeder reactor

    International Nuclear Information System (INIS)

    The purpose of the promotion policy for energy research and development base construction plan (priority facility) of the Japanese government in FY2009 is 'to construct in Tsuruga City the research and development base for plant operation technology for the practical use of fast breeder reactor where researchers in and out of Japan gather, and to contribute to the development and revitalization of the region as the base with international characteristics.' In conformity to this purpose, the Japan Atomic Energy Agency built 'sodium engineering research facilities' in Tsuruga. This paper describes the design, fabrication, and installation of interior equipment that were carried out by Kawasaki Heavy Industries. 'Sodium engineering research facilities' are the test and research facilities to conduct research and development related to sodium, while reflecting the experiences of operation and maintenance of 'Monju,' which aims at the commercialization of fast reactor. The facilities specialize in the handling technology of sodium to meet the needs in and out of Japan, and were completed in June 2015. The facilities consist of six units including tank-loop test equipment, mini-loop test equipment, sodium purification and supply equipment, etc. For the tank-loop test equipment, a sodium transfer test of about 5.5 tons, and a subsequent comprehensive function test using sodium are scheduled. (A.O.)

  13. Engineering design and development for prototype fast breeder reactor (PFBR) shielding experiments at Apsara

    International Nuclear Information System (INIS)

    Prototype fast breeder reactor (PFBR) houses radial shields inside the reactor vessel which consists of many layers of steel and borated graphite within sodium coolant so as to reduce the neutron flux impingement on Intermediate Heat Exchanger (IHX) (also located inside the reactor vessel) to an acceptable limit. In order to cross check the uncertainties involved in theoretical shielding calculations and neutron cross-section data used, IGCAR proposed to carry out various shielding experiments at Apsara reactor to simulate the theoretical shielding configuration. The experiments would also provide bias factors for detailed shielding design calculations. The shielding experiments were planned to be carried out at Apsara shielding corner with reactor core brought to C-dash (C) position. The neutron flux intensity in the shielding corner was inadequate for the purpose of carrying out experiments. Hence the neutron flux level was enhanced to the order of 1010 n/cm2/s by replacing the water column between the core edge and SS liner of Apsara pool on the shielding corner side with an air filled leak tight aluminium box. The fuel loading in the reactor core was also modified to increase neutron flux intensity towards aluminium box. The neutron flux emerging out of the pool into the shielding corner is essentially a thermal neutron spectrum, which was converted into a typical fast reactor leakage neutron spectrum with the help of converter assemblies (CAs ). The converter assemblies were made of depleted uranium and the assemblies were installed on a CA trolley. The CA trolley was positioned outside Apsara pool in the shielding corner. The models of proposed shields manufactured from various shielding materials viz. sodium, steel, borated graphite and boron carbide were installed on a shield model (SM) trolley. The SM trolley was positioned behind CA trolley. Shield models had provisions for irradiating in any foils which were used for measuring the neutron attenuation

  14. Design, implementation and cost-benefit analysis of a dynamic testing program in the Experimental Breeder Reactor-II

    International Nuclear Information System (INIS)

    Dynamic tests have been performed for many years in commercial pressurized and boiling water reactors. The purpose of this study was to evaluate the technological and economical feasibility of extending the current light water reactor testing procedures to both present and future liquid metal fast breeder reactors. A 38 node linearized, lumped parameter, EBR-II system model was developed. This model was analyzed to obtain the predicted system time and frequency response for reactivity perturbations, intermediate heat exchanger secondary inlet sodium temperature perturbation frequency response, and various system nodal frequency response sensitivities

  15. Summary of several hydraulic tests in support of the light water breeder reactor design (LWBR development program)

    International Nuclear Information System (INIS)

    As part of the Light Water Breeder Reactor development program, hydraulic tests of reactor components were performed. This report presents the results of several of those tests performed for components which are somewhat unique in their application to a pressurized water reactor design. The components tested include: triplate orifices used for flow distribution purposes, multiventuri type flowmeters, tight lattice triangular pitch rod support grids, fuel rod end support plates, and the balance piston which is a major component of the movable fuel balancing system. Test results include component pressure loss coefficients, flowmeter coefficients and fuel rod region pressure drop characteristics

  16. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  17. Problems of nuclear reactor safety. Vol. 2

    International Nuclear Information System (INIS)

    Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

  18. Nuclear problems of power reactors safety

    International Nuclear Information System (INIS)

    The objective of this presentation is to emphasize the conditions that would be of high importance in safety analyses concerned first of all with reactor core. It describes reactor kinetics processes in the core and build up of fission products, classification of reactor accidents related to the core, risk estimation and includes a list of importance reactor accidents

  19. Theoretical and experimental studies of non-linear structural dynamics of fast breeder reactor fuel elements

    International Nuclear Information System (INIS)

    Descriptions are presented of theoretical and experimental studies of the deformation behaviour of fast-breeder fuel elements as a consequence of extreme impulsive stresses produced by an incident. The starting point for the studies is the assumption that local disturbances in a fuel element have resulted in a thermal interaction between fuel and sodium and in a corresponding increase in pressure. On the basis of the current state of knowledge, the possibility cannot be ruled out that this pressure build-up may lead to the bursting of the fuel-element wrapper, to the propagation of pressure in the core, and to coherent structural movements and deformations. A physical model is established for the calculation of the dynamic response of elastic-plastic beam systems, and the differential equations of p motion for the discrete equivalent system are derived with the aid of D'Alembert's principle. On this basis and with the aid of a semi-empirical pin-bundle model, an appropriate computer program allows a static and dynamic analysis to be obtained for a complete fuel element. In the experimental part of the study, a description is given of static and impulsive loading tests on 1:1 SNR-like fuel-element models. Making use of measured impact forces and of known material characteristics, it was possible to a large extent for the experiments to be reproduced by calculations. In agreement with existing experience from explosion experiments on 1:1 core models, the results (of relevance for fast-breeder safety and in particular the SNR-300) show that only local limited deformations occur and that the compact fuel-element and core structure constitutes an effective inherent barrier in the presence of extreme incident stresses. (author)

  20. Safety Management at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Adequate safety measures and precautions, which follow relevant safety standards and procedures, should be in place so that personnel safety is assured. Nevertheless, the public, visitor, contractor or anyone who wishes to enter or be in the reactor building should be well informed with the safety measures applied. Furthermore, these same elements of safety are also applied to other irradiation facilities within the premises of Nuclear Malaysia. This paper will describes and explains current safety management system being enforced especially in the TRIGA PUSPATI Reactor (RTP) namely radiation monitoring system, safety equipment, safe work instruction, and interconnected internal and external health, safety and security related departments. (author)

  1. Fast breeder reactor blanket management: comparison of LMFBR and GCFR blankets

    International Nuclear Information System (INIS)

    The economic performance of the fast breeder reactor blanket, considering different fuel management schemes was studied. To perform this, the investigation started with a standard reactor physics calculation. Then, two economic models for evaluation of the economic performance of the radial blanket were developed. These models formed the basis of a computer code, ECOBLAN, which computes the net economic gain and the levelized fuel cost due to the radial blanket. The net gain in terms of dollars and $/kgHM-y and the levelized fuel cost in mills/kWhe were obtained as a function of blanket thickness and a residence time of the fuel in the blanket. A LMFBR and a GCFR were the reactor models considered in this study. The optimum radial blanket of a GCFR consists of two rows, that of a LMFBR consists of three rows. Regarding the different fuel management schemes, the fixed blanket was found to be more favorable than reshuffled blanket. Out-in and in-out reshuffled blanket offer almost the same net gain. In all the cases, the burnup calculated for the fuel was found to be less than the acceptable limit. There is an optimum residence time for the fuel in the blanket which depends on the position of the fuel in the blanket and the fuel management scheme studied. As expected, except for very short residence times (less than 2.5 years), the radial blanket is a net income producer. There is no significant difference between the economic performance of the blanket of a LMFBR and a GCFR

  2. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  3. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design

  4. Web-enabled work permit system for fast breeder test reactor

    International Nuclear Information System (INIS)

    The objective of this project is to computerize and web-enable the Work Permit System for the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam. The existing Work Permit System at FBTR was studied in detail. Since all the formalities were paper-based, the risk of human error in scrutinizing all permits before reactor start-up was high. Compilation of reports (daily, monthly, yearly etc.) was tedious. The work permit system was therefore automated in order to enable the operation group manage the maintenance work carried out in the plant systematically with entries. The entire project was classified into five permit modules -maintenance, transfer, return, cancellation and reissue. Each module takes care of the entry and maintenance of data in their respective fields in their respective tables. The user is also provided with an option to take a hard copy of the report of his/her choice. A client/server based system was designed to web-enable the entire project. The server program was designed using VB 6.0 as the front-end and MS Access database as the back end to store the data. The client software was developed using Active Server Pages and published using personal web server in the Intranet. A number of administrative tools have been incorporated in the software to ensure access security and integrity of the database. An online help feature with search facilities was added to the software. The work permit system software is now already being used at FBTR and has been deemed to be an invaluable aid in empowering the availability of the reactor and determining the performance history of the equipment. (author)

  5. Safety analysis for non-power reactors

    International Nuclear Information System (INIS)

    Non-power reactors have been operating in Canada since 1945, with NRU (National Research Universal, 1957) being the oldest operating non-power reactor. Presently, there are five generic 'types' of non-power reactors: NRU, ZED-2, SLOWPOKE, MNR and MAPLE, the latter undergoing commissioning as the MDS Medical Isotope Reactor. These reactors range in thermal power from 200 Watts to more than 100 MW. Other non-power reactors are likely to be built for new applications and to replace older reactors. The uniqueness of each reactor, the wide range of power levels and the evolution of safety philosophy over time have lead to non-uniform practices for safety analysis. This non-uniformity may be a problem for the preparation by the licensee and review by the regulator of the safety analysis report required for licensing of the reactor facility. Clearly, there is no universally applicable practice, while at the same time, expectations for safety analyses have evolved in order to demonstrate higher levels of overall safety. This paper examines a new 'graded approach' to preparing the safety analysis report for reactors of diverse features but with a common standard of safety. It discusses necessary content, methods and the training and qualification of the safety analyst. (author)

  6. Data management for the Clinch River Breeder Reactor Plant Project by use of document status and hold systems

    International Nuclear Information System (INIS)

    This paper describes the development, framework, and scope of the Document Status System and the Document Hold System for the Clinch River Breeder Reactor Plant Project. It shows how data are generated at five locations and transmitted to a central computer for processing and storage. The resulting computerized data bank provides reports needed to perform day-to-day management and engineering planning. Those reports also partially satisfy the requirements of the Project's Quality Assurance Program

  7. Breeder Reprocessing Engineering Test

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, C.A.; Meacham, S.A.

    1984-01-01

    The Breeder Reprocessing Engineering Test (BRET) is a developmental activity of the US Department of Energy to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the Fast Flux Test Facility (FFTF). It will be installed in the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site near Richland, Washington, The major objectives of BRET are: (1) close the US breeder fuel cycle; (2) develop and demonstrate reprocessing technology and systems for breeder fuel; (3) provide an integrated test of breeder reactor fuel cycle technology - rprocessing, safeguards, and waste management. BRET is a joint effort between the Westinghouse Hanford Company and Oak Ridge National Laboratory. 3 references, 2 figures.

  8. Techniques developed to evaluate the fracture toughness offast breeder reactor duct

    International Nuclear Information System (INIS)

    Large changes in strength and ductility of metals after irradiation are known to occur. The fracture toughness of irradiated metals, which is related to the combined strength and ductility of a material, may be significantly reduced and the potential for unstable crack extension increased. Therefore, the resistance of cladding and duct materials to fracture after exposure to fast neutron environments is of concern. Existing Type 316 stainless steel irradiated ducts are relatively thin and since this material retains substantial ductility, even after irradiation, the fracture behavior of the duct material cannot be analyzed by linear elastic fracture mechanics techniques. Instead, the multispecimen R-curve method and J-integral analysis were used to develop an experimental approach to evaluate the fracture toughness of thin breeder reactor duct materials irradiated at elevated temperatures. Alloy A-286 was chosen for these experiments because the alloy exhibits elastic/plastic behavior and the fracture toughness data of thicker (12 mm) specimens were available for comparison. Technical problems associated with specimen buckling and remote handling were treated in this work. The results are discussed in terms of thickness criterion for plane strain

  9. Reactivity control capability of fuel-salt processing system in a molten-salt breeder reactor

    International Nuclear Information System (INIS)

    An evaluation is made of the reactivity control capability of the fuel processing system (FPS) in a molten-salt breeder reactor. The principal functions required of the FPS are: (a) Isolation of 233Pa from regions of high neutron flux during its decay to 233U, and (b) the removal of fission products from the system. The FPS can very usefully serve also to control the primary system reactivity by appropriately utilizing its function of extracting uranium and reconstituting the fuel contained in the salt. The principles of operation are quite similar to the chemical shim control system currently installed in PWR's whereby the core reactivity, affected by changes in the moderator temperature, fuel burnup and transient Xe, is adjusted by regulating the concentration of boric acid introducted in the moderator as neutron absorber. The present study examines the capability of the FPS to follow transient Xe as in PWR's, and proves that the FPS should effectively serve as a system for adjusting not only long-term changes in reactivity but also short-term transient variations without any accompanying difficulties foreseen in operation. (auth.)

  10. Blowdown transient for sodium-steam water SG for prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lele, H.G.; Srivastava, A.; Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K. [Reactor Safety Div., Bhabha Atomic Research Centre, Tromblay (India); Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Associate Director, Reactor Group, Chennai (India)

    2001-07-01

    Prototype Fast Breeder Test Reactor (PFBR) Steam Generator is once through steam generator in which water flows from bottom to top in 547 tubes, changing its state from highly subcooled to superheated state as it receives heat from sodium flowing from top to bottom in the shell side. Depressurization of steam generator from the dump valve provided at bottom is protective action. It prevents further possibility of water steam leak into sodium and subsequent sodium - water reaction. To perform depressurization transient analysis of PFBR appropriate thermal hydraulic modeling of SG is essential. Correct thermal hydraulic modelling needs simulation of sodium system, steam water system with different states from highly subcooled to superheated, coupling between sodium and steam-water system, SG tube and shell and different valve action. The computer code DPPFBR is developed with capability to simulate all these systems and phenomena encountered during transient. Different models of the code have been validated and code has been used for analysing depressurization transient. This paper describes various models used in the code and results of analysis for typical scenario. (author)

  11. Effect of yttrium additions on void swelling in Liquid Metal Fast Breeder Reactor candidate cladding alloys

    International Nuclear Information System (INIS)

    Candidate Liquid Metal Fast Breeder Reactor cladding alloys AL1 (Fe-26% Ni-9% Cr) and AL2 (Fe-35% Ni-12% Cr) without and with the addition of 0.1% yttrium were bombarded by 4 MeV56Fe2+ ions without and with simultaneous bombardment by 0.4 MeV 4He+ ions. These bombardments were conducted at various irradiation temperatures to determine the effect of yttrium on void swelling. The addition of yttrium decreased peak swelling for 4 MeV 56Fe2+ ion bombarded AL1 and AL2 by 28% and 20%, respectively. In all cases where similar sample comparisons were made (i.e., undoped with undoped and doped with doped) and where bombardment conditions were similar (i.e., single with single beam and dual with dual beam), AL1 showed less peak swelling than did AL2. Simultaneously implanting helium during heavy-ion bombardment increased peak swelling in undoped and doped AL1 by factors of 2.3 and 2.6, respectively

  12. Engineering development studies for molten-salt breeder reactor processing No. 18

    International Nuclear Information System (INIS)

    A water--mercury system was used to study the effect of geometric variations on mass transfer rates in rectangular contractors similar to those proposed for the molten-salt breeder reactor (MSBR) fuel reprocessing scheme. Since mass transfer rates were not accurately predicted by the Lewis correlation, other correlations were investigated. A correlation which was found to fit the experimental results is given. Mass transfer rates are being measured in a fluoride salt--bismuth contactor. Experimental results indicate that the mass transfer rates in the salt--bismuth system fall between the Lewis correlation and the modified correlation given above. Autoresistance heating tests were continued in the fluorinator mock-up using LiF--BeF2--ThF4 (72-16-12 mole percent) salt. The equipment was returned to operating condition, and five experiments were run. Although correct steady-state operation was not achieved, the results were encouraging. A two-dimensional electrical analog was constructed to study current flow through the electrode sidearm and other critical areas of the test vessel. These studies indicate that no regions of abnormally high current density existed in the first nine runs with the present autoresistance heating equipment. Localized heating had previously been the suspected cause for the failure to achieve proper operation of this equipment. (U.S.)

  13. Development of an ISI robot for the fast breeder reactor MONJU primary heat transfer system piping

    International Nuclear Information System (INIS)

    This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire-type ultrasonic sensor for volumetric tests at high temperature (atmosphere, 55degC; piping surface, 80degC) and radiation exposure condition (dose rate, 10 mGy/h; piping surface dose rate, 15 mGy/h). An inspection robot using a new tire type for the ultrasonic testing sensor and a new control method was developed. A signal-to-noise ratio S/N over 2 was obtained during the functional test for a calibration defect with a depth of 50%t (from the tube wall thickness). In the automatic inspection test, an EDM slit with a depth of 9% from the pipe thickness was detectable and with an S/N ratio = 4.0 (12.0 dB). (author)

  14. Calculations of sodium aerosol concentrations at breeder reactor air intake ports

    International Nuclear Information System (INIS)

    This report describes the methodology used and results obtained in efforts to estimate the sodium aerosol concentrations at air intake ports of a liquid-metal cooled, fast-breeder nuclear reactor. A range of wind speeds from 2 to 10 m/s is assumed, and an effort is made to include building wake effects which in many cases dominate the dispersal of aerosols near buildings. For relatively small release rates on the order of 1 to 10 kg/s, it is suggested that the plume rise will be small and that estimates of aerosol concentrations may be derived using the methodology of Wilson and Britter (1982), which describes releases from surface vents. For more acute releases with release rates on the order of 100 kg/s, much higher release velocities are expected, and plume rise must be considered. Both momentum-driven and density-driven plume rise are considered. An effective increase in release height is computed using the Split-H methodology with a parameterization suggested by Ramsdell (1983), and the release source strength was transformed to rooftop level. Evaluation of the acute release aerosol concentration was then based on the methodology for releases from a surface release of this transformed source strength

  15. Tridimensional ultrasonic images analysis for the in service inspection of fast breeder reactors

    International Nuclear Information System (INIS)

    Tridimensional image analysis provides a set of methods for the intelligent extraction of information in order to visualize, recognize or inspect objects in volumetric images. In this field of research, we are interested in algorithmic and methodological aspects to extract surface visual information embedded in volume ultrasonic images. The aim is to help a non-acoustician operator, possibly the system itself, to inspect surfaces of vessel and internals in Fast Breeder Reactors (FBR). Those surfaces are immersed in liquid metal, what justifies the ultrasonic technology choice. We expose firstly a state of the art on the visualization of volume ultrasonic images, the methods of noise analysis, the geometrical modelling for surface analysis and finally curves and surfaces matching. These four points are then inserted in a global analysis strategy that relies on an acoustical analysis (echoes recognition), an object analysis (object recognition and reconstruction) and a surface analysis (surface defects detection). Few literature can be found on ultrasonic echoes recognition through image analysis. We suggest an original method that can be generalized to all images with structured and non-structured noise. From a technical point of view, this methodology applied to echoes recognition turns out to be a cooperative approach between morphological mathematics and snakes (active contours). An entropy maximization technique is required for volumetric data binarization. (author)

  16. Clinch River Breeder Reactor environmental effects: general water-side corrosion

    International Nuclear Information System (INIS)

    Studies are described of the general corrosion of 21/4 Cr--1 Mo steel in pure superheated steam, in impure superheated and saturated steam, and under nucleate boiling conditions. The test parameters were selected to provide information relevant to the use of this steel for the Clinch River Breeder Reactor superheaters and evaporators. The oxidation rate of 21/4 Cr--1 Mo steel in superheated steam was measured under heat transfer conditions at 510 to 5400C (950 to 10050F), and was approximately 11/2 times that measured under isothermal conditions. Extensive general attack of stressed 21/4 Cr--1 Mo steel specimens occurred in cyclic tests in superheated and saturated steam with chloride and oxygen additions, although no cracking or localized attack was observed. Considerably less attack occurred under superheat conditions or in the absence of oxygen. Tests under nucleate boiling conditions were operated to evaluate crevice effects associated with porous films on heat transfer surfaces. Significant crevice corrosion was produced in water containing 10 ppm chloride; a heavier but more general attack occurred in treated cooling tower water

  17. The development of a cavitation free sodium pump for the breeder reactor

    International Nuclear Information System (INIS)

    The sodium pumps for a liquid metal fast breeder reactor must be designed for exceptionally high reliability and long life. The principal adverse factor which tends to limit the primary pump life is cavitation which becomes potentially severe under off-design flow conditions caused by the requirement of two loop operations which resulted in a large operating flow range. This problem prompted an extensive study which included experimental investigations of scaled down and full size pumps. The investigations involved visual observations, acoustic signature recordings, and physical characteristic measurements of the model and full size impellers. The blade configuration of the model was modified several times. After each modification intensive testing was conducted with feedback to established design criteria. The results obtained from the final configuration showed excellent cavitation performance. This configuration was then machined on the full scale impeller and tested. The results confirmed acceptable performance in the entire range of operating conditions. This paper describes the test facilities erected for this study, discusses the experimental techniques employed, and presents the experimental techniques employed, and presents a sample of the experimental results

  18. Leakage limits for inflatable seals of sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • All possible types/modes of gas escape covered. • Limits include simultaneous contributions from bypass and permeation leakage modes. • Leakage of radioactive cover gas with fission products assumed. • Possibility of sodium frost deposition in sealed gap considered. • Cover gas activity decay during fuel handling and relative importance of types/modes of leakage considered for realistic results and simpler seal design. -- Abstract: Estimation and stipulation of allowable leakage for inflatable seals of 500 MWe Prototype Fast Breeder Reactor is depicted. Leakage limits are specified using a conservative approach, which assumes escape of radioactive cover gas with fission products across the seals in bypass and permeation modes and possibility of sodium frost deposition in sealed gaps because of permeation leakage of inflation gas. Procedures to arrive at the allowable leakages of argon cover gas (normal-operation/fuel-handling: 10−3/10−2 scc/s/m length of seal) and argon inflation gas (10−3 scc/s/m length of seal) is described

  19. Properties of structural materials for sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    For the selection of the structural materials for superheaters and reheaters of sodium-cooled fast breeder reactors, it is important to grasp the change of strength due to the complex change of material properties, such as the combination of surface corrosion, the decarbonization of 2-1/4 Cr-1Mo steel, the carburization of austenitic stainless steel, the structural change due to heating history, etc. in sodium environment. The stress corrosion cracking of austenitic materials in water must be studied also. The materials taken up in this paper are austenitic stainless steels such as SUS 304, SUS 316, SUS 321, and SUS 347, iron-based superalloy Incoloy 800, and ferritic alloy steel 2-1/4Cr-1Mo steel. The data on the above described properties of the materials are given. Also the tensile strength, creep rupture and fatigue characteristics of the parent materials in the amount of corrosion in sodium. The strength of ferritic alloy steel is lowered owing to the decarbonization in sodium, but the change of strength due to carburization was not observed. There is some possibility that the unstabilized steels such as SUS 304 and SUS 316 become sensitive to stress corrosion cracking, and the stabilized steels such as SUS 321 and SUS 347 become sensitive to it in long hour heating. The tensile strength of welded joints is almost same as that of parent materials, but the elongation decreases by about 10%. (Kako, I.)

  20. Performance characterization of geopolymer composites for hot sodium exposed sacrificial layer in fast breeder reactors

    International Nuclear Information System (INIS)

    Highlights: • Performance evaluation of geopolymers subjected to hot liquid sodium is performed. • Apart from mechanical properties, micro-analytical techniques are used for material characterization. • The geopolymer composite showed comparatively lesser damage than conventional cement composites. • Geopolymer technology can emerge as a new choice for sacrificial layer in SCFBRs. - Abstract: A sacrificial layer of concrete is used in sodium cooled fast breeder reactors (SCFBRs) to mitigate thermo-chemical effect of accidentally spilled sodium at and above 550 °C on structural concrete. Performance of this layer is governed by thermo-chemical stability of the ingredients of sacrificial layer concrete. Concrete with limestone aggregate is generally used as a sacrificial layer. Conventional cement based systems exhibit instability in hot liquid sodium environment. Geo-polymer composites are well known to perform excellently at elevated temperatures compared to conventional cement systems. This paper discusses performance of such composites subjected to exposure of hot liquid sodium in air. The investigation includes comprehensive evaluation of various geo-polymer composites before any exposure, after heating to 550 °C in air, and after immersing in hot liquid sodium initially heated to 550 °C in air. Results from the current study indicate that hot liquid sodium produces less damage to geopolymer composites than to the existing conventional cement based system. Hence, the geopolymer technology has potential application in mitigating the degrading effects of sodium fires and can emerge as a new choice for sodium exposed sacrificial layer in SCFBRs

  1. Pulsed Nd-YAG laser welding of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    End plug welding of Prototype Fast Breeder Reactor (PFBR) fuel elements involves welding of fully Austenitic Stainless Steel (ASS) of grade D9 clad tube with 316M end plug. Pulsed Gas Tungsten Arc Welding (GTAW) is being used for the production of PFBR fuel elements at Advanced Fuel Fabrication Facility (AFFF). GTAW is an established process for end plug welding and hence adopted by many countries. GTAW has got certain limitations like heat input, arc gap sensitivity and certain sporadic defects like tungsten inclusion. Experiments have been carried out at AFFF to use Laser Beam Welding (LBW) technique as LBW offers a number of advantages over the former process. This report mainly deals with the optimization of laser parameters for welding of PFBR fuel elements. To facilitate pulsed Nd-YAG laser spot welding, parameters like peak power, pulse duration, pulse energy, frequency and defocusing of laser beam on to the work piece have been optimized. On the basis of penetration requirement laser welding parameters have been optimized. (author)

  2. A study of parameters on marking of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor Fuel (PFBR) elements are identified with a permanent unique marking. Identification of the fuel elements is very much necessary for traceability during initial fabrication as well as for post irradiation examination. Marking on fuel element has to be permanent and capable of being identified after irradiation. Laser marking is a relatively new method as compared to other marking technologies such as ink marking, mechanical engraving and electro chemical methods. It is used for the product identification and traceability during its service life. Laser marking has many advantages compared to other conventional marking. In laser marking process, mark quality is a very important factor, which depends on so many variables like input current, pulse frequency, marking speed and number of passes. The influence of the pulse frequency and the speed of travel of the laser beam on the mark depth and width have been studied in this paper. An optical microscope, scanning electron microscope were used to measure the effects of pulse frequency on the mark depth and width. It has been found that the mark depth and width depend on the interaction process of the laser beam and the material, which was influenced by the pulse frequency. Micro hardness testing is carried out to report Heat Affected Zone (HAZ) variation with parameters. Marking speed and input current selected for suitable depth and width were mentioned in the present study. (author)

  3. Seismic analysis of primary sodium pump of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    This paper deals with seismic analysis of primary sodium pump of 500 MWe Prototype Fast Breeder Reactor (PFBR) by time history analysis. The associated fluid mechanics equations for determining stiffness coefficients for hydrostatic bearing (HSB) are formulated and solved by integrating with the governing structural dynamics equations. Thus, the stiffness coefficients of HSB are computed as function of eccentricity, speed and angular orientation of shaft at every solution time step. The maximum stress intensity values in the flywheel-motor support shell, pump shell and shaft are insignificant (<20 MPa). Hence, the structural integrity is assured for pump parts. The seizure of pump is found to be a critical issue under seismic events. The eccentricity of 60 μ experienced by the shaft during normal operation, increases to 250 μ under SSE. Since it is < 400 μ, the radial clearance, there is no problem of pump seizure. However, at speeds lower than 265 rpm, the eccentricity of shaft is higher than radial clearance and hence shaft can impact on the shell. Under OBE, up to 20 % speed (∼120 rpm), there is no impact. The peak impact force experienced by shaft and shell at HSB at 235 rpm under SSE is < 3 t, which may not cause pump seizure. This needs to be confirmed by experiments. (author)

  4. Fabrication and loading of fuel rods for the Light Water Breeder Reactor (LWBR Development Program)

    International Nuclear Information System (INIS)

    The fabrication and inspection operations used for the manufacture of approximately 24,000 fuel rods for the Light Water Breeder Reactor are described in detail. This report also describes the development work to establish the fabrication procedures and investigations undertaken to solve problems encountered during manufacturing. The approximately 10 foot long LWBR fuel rods were made in four outside diameters ranging from 0.306 inch (seed) to 0.832 inch (reflector). Each rod was fabricated by sealing cylindrical oxide fuel pellets (ThO2-U233O2), into Zircaloy seamless tubes by welding Zircaloy enclosures at the ends. The special inspections performed to assure a high quality product meeting all design requirements are described. These inspections included weld radiography and ultrasonic inspection, in-motion radiography to evaluate internal dimensions and pellet integrity, helium leak testing, corrosion testing, and detection of surface contamination. The facilities designed and built for this fabrication effort are described and the resultant manufacturing yields are presented. 13 refs., 42 figs., 20 tabs

  5. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  6. Blowdown transient for sodium-steam water SG for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Test Reactor (PFBR) Steam Generator is once through steam generator in which water flows from bottom to top in 547 tubes, changing its state from highly subcooled to superheated state as it receives heat from sodium flowing from top to bottom in the shell side. Depressurization of steam generator from the dump valve provided at bottom is protective action. It prevents further possibility of water steam leak into sodium and subsequent sodium - water reaction. To perform depressurization transient analysis of PFBR appropriate thermal hydraulic modeling of SG is essential. Correct thermal hydraulic modelling needs simulation of sodium system, steam water system with different states from highly subcooled to superheated, coupling between sodium and steam-water system, SG tube and shell and different valve action. The computer code DPPFBR is developed with capability to simulate all these systems and phenomena encountered during transient. Different models of the code have been validated and code has been used for analysing depressurization transient. This paper describes various models used in the code and results of analysis for typical scenario. (author)

  7. Optimal measurement uncertainties for materials accounting in a fast breeder reactor spent-fuel reprocessing plant

    International Nuclear Information System (INIS)

    Optimization techniques are used to calculate measurement uncertainties for materials accountability instruments in a fast breeder reactor spent-fuel reprocessing plant. Optimal measurement uncertainties are calculated so that performance goals for detecting materials loss are achieved while minimizing the total instrument development cost. Improved materials accounting in the chemical separations process (111 kg Pu/day) to meet 8-kg plutonium abrupt (1 day) and 40-kg plutonium protracted (6 months) loss-detection goals requires: process tank volume and concentration measurements having precisions less than or equal to 1%; accountability and plutonium sample tank volume measurements having precisions less than or equal to 0.3%, short-term correlated errors less than or equal to 0.04%, and long-term correlated errors less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having precisions less than or equal to 0.4%, short-term correlated errors less than or equal to 0.1%, and long-term correlated errors less than or equal to 0.05%

  8. Safety of research reactors - A regulator's perspective

    International Nuclear Information System (INIS)

    Due to historical reasons research reactors have received less regulatory attention in the world than nuclear power plants. This has given rise to several safety issues which, if not addressed immediately, may result in an undesirable situation. However, in Pakistan, research reactors and power reactors have received due attention from the regulatory authority. The Pakistan Research Reactor-1 has been under regulatory surveillance since 1965, the year of its commissioning. The second reactor has also undergone all the safety reviews and checks mandated by the licensing procedures. A brief description of the regulatory framework, the several safety reviews carried out have been briefly described in this paper. Significant activities of the regulatory authority have also been described in verifying the safety of research reactors in Pakistan along with the future activities. The views of the Pakistani regulatory authority on the specific issues identified by the IAEA have been presented along with specific recommendations to the IAEA. We are of the opinion that there are more Member States operating nuclear research reactors than nuclear power plants. Therefore, there should be more emphasis on the research reactor safety, which somehow has not been the case. In several recommendations made to the IAEA on the specific safety issues the emphasis has been, in general, to have a similar documentation and approach for maintaining and verifying operational safety at research reactors as is currently available for nuclear power reactors and may be planned for nuclear fuel cycle facilities. (author)

  9. IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

  10. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    Energy Technology Data Exchange (ETDEWEB)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  11. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    International Nuclear Information System (INIS)

    The technology of breeding 233U from 232Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program

  12. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  13. Two decades of experience with steam-water chemistry maintenance of fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) at Kalpakkam is a 40 MWt, loop type, sodium cooled fast reactor. The fission heat generated in the core is extracted by primary sodium circuit and the thermal energy is transferred to non-radioactive liquid sodium in the secondary circuit which in turn, heats Once Through-type shell and tube counter current Steam Generator (OTSG) for producing super heated steam at 480 °C and 125 kg/cm2. This secondary circuit is provided to avoid the ingress of hydrogenous materials and pressure surges reaching the core in the event of SG tube leak. Corrosion related problems are very less in the sodium circuits due to the absence of electrochemical reaction. The OTSG consists of four modules each of 12.5 MWt rating. OTSG was chosen due to its higher thermal efficiency and lesser inventory of steam/water in OTSG as it reduces the severity of sodium-water reaction, in case of tube leak. From the point of view of corrosion and deposition, the chemistry specifications are more stringent for OTSG than those of drum type boilers because 100 % conversion of feed water into steam takes place in OTSG. The chemistry requirements are achieved by providing ion exchange resin based online condensate polishing to remove ionic and suspended impurities. Dissolved Oxygen and pH are maintained by all volatile treatment (AVT) using hydrazine and ammonia respectively. Being a test reactor, a dump condenser with 100 % steam dump facility with cupro-nickel tubes is available for uninterrupted reactor operation during the non-availability of turbine. Regenerative feed heating by the exhausted steam from the turbine is also available to stage heaters and deaerator. Efficient water chemistry control plays important role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. This paper describes the operational difficulties such as premature exhaustion of CPU, impurity pick up from the system, silica excursion

  14. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%

  15. Interaction of CREDO [Centralized Reliability Data Organization] with the EBR-II [Experimental Breeder Reactor II] PRA [probabilistic risk assessment] development

    International Nuclear Information System (INIS)

    The National Academy of Sciences review of US Department of Energy (DOE) class 1 reactors recommended that the Experimental Breeder Reactor II (EBR-II), operated by Argonne National Laboratory (ANL), develop a level 1 probabilistic risk assessment (PRA) and make provisions for level 2 and level 3 PRAs based on the results of the level 1 PRA. The PRA analysis group at ANL will utilize the Centralized Reliability Data Organization (CREDO) at Oak Ridge National Laboratory to support the PRA data needs. CREDO contains many years of empirical liquid-metal reactor component data from EBR-II. CREDO is a mutual data- and cost-sharing system sponsored by DOE and the Power Reactor and Nuclear Fuels Development Corporation of Japan. CREDO is a component based data system; data are collected on components that are liquid-metal specific, associated with a liquid-metal environment, contained in systems that interface with liquid-metal environments, or are safety related for use in reliability/availability/maintainability (RAM) analyses of advanced reactors. The links between the EBR-II PRA development effort and the CREDO data collection at EBR-II extend beyond the sharing of data. The PRA provides a measure of the relative contribution to risk of the various components. This information can be used to prioritize future CREDO data collection activities at EBR-II and other sites

  16. Potential of duplex fuel in prebreeder, breeder, and power reactor designs: tests and analyses (AWBA Development Program)

    International Nuclear Information System (INIS)

    Dual region fuel pellets, called duplex pellets, are comprised of an outer annular region of relatively high uranium fuel enrichment and a center pellet of fertile material with no enrichment. UO2 and ThO2 are the fissile and fertile materials of interest. Both prebreeders and breeders are discussed as are the performance advantages of duplex pellets over solid pellets in these two pressurized water reactor types. Advantages of duplex pellets for commercial reactor fuel rods are also discussed. Both irradiation test data and analytical results are used in comparisons. Manufacturing of duplex fuel is discussed

  17. Uncertainty in the breeding ratio of a large liquid-metal fast breeder reactor: theory and results

    International Nuclear Information System (INIS)

    Using an extensive data base of sensitivities and evaluated covariances, this work incorporates 11 fast-reactor benchmark experiments and 2 neutron-field benchmark experiments into the adjustment of a 26-group cross-section library based primarily on Evaluated Nuclear Data File (ENDF)/B-IV. The adjustments of the group cross sections are examined in detail. The results of the adjustment are applied to the determination of the uncertainties in the multiplication factor and in the breeding ratio of a large liquid-metal fast breeder reactor design model fixed by the Lartge Core Code Evaluation Working Group. 71 refs

  18. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  19. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    Energy Technology Data Exchange (ETDEWEB)

    None

    1983-10-01

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  20. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  1. Zeolite membranes and palladium membrane reactor for tritium extraction from the breeder blankets of ITER and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Demange, D., E-mail: david.demange@kit.edu; Borisevich, O.; Gramlich, N.; Wagner, R.; Welte, S.

    2013-10-15

    Highlights: • We present a new concept to recover tritium from the helium in breeder blankets. • Zeolite membranes are fully tritium compatible and can pre-concentrate tritiated molecules. • PERMCAT catalytic membrane reactor recovers tritium to be reused in the fuel cycle. -- Abstract: While the tritium technology for the inner DT fuel cycle of fusion reactors shall be demonstrated in ITER, the tritium management in the breeder blanket remains very challenging. Most of the process options rely on ad(b)sorption/desorption cycles, using dedicated packed beds to handle separately the molecular and oxide forms of tritium. This approach seems satisfactory for ITER, but seems difficult to scale up to DEMO. The alternative use of a catalytic membrane reactor in combination with inorganic membranes would simplify and improve the overall tritium management. Zeolite membranes should enable in a single step the pre-concentration of all tritiated species. This tritium enriched stream could be afterwards processed using PERMCAT (catalytic Pd-based membrane reactor) to finally recover the tritium in its pure molecular form. This paper discusses at the conceptual level such approach. The latest experimental results on zeolite membrane and multi-tube PERMCAT reactor are presented. Next R and D activities for technical scale demonstrations and refined simulation tools are proposed to finally estimate the sizes of the components to be operated in ITER and DEMO.

  2. The role and problems of the breeder

    International Nuclear Information System (INIS)

    World uranium resources are discussed and it is concluded that the period of availability of uranium for use in the present type of nuclear power stations is not much greater than that of oil. The neutron economies of fast and thermal reactors are compared, and the advantages of the breeder for the world uranium economy are demonstrated. The main impediments to the use of the fast breeder are considerations of safety, public acceptance and economics. Fast reactor safety is discussed and the health hazards and possible mis-use of plutonium for terrorism and weapons proliferation are considered. It is widely accepted that the U.K. cannot economically justify the development of the breeder alone and is likely to choose to co-operate with Western Europe. A public enquiry in the U.K. seems certain and would be welcomed by the nuclear industry. (author)

  3. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  4. Euratom research supporting reactor safety

    International Nuclear Information System (INIS)

    This paper focuses on the way the Fifth Framework Programme (FP5) of the European Atomic Energy Community (Euratom) (1998-2002) contributed to the overall effort supporting LWR safety and on the prospects of FP6 (2002-2006) to achieve the additional goal of contributing to the creation of the European Research Area (ERA). The objectives of the FP5 Community research are discussed, topic per topic, with reference to the relevant projects contributing to their achievement while emphasising the role and expectations of the end-users. 71 research projects have been carried out in FP5 with a total cost of 85.4 million Euro, out of which 43.5 million Euro are contributed by the EU. Economic aspects and needs for nuclear knowledge management throughout the EU-25 countries are briefly discussed as well as future Euratom research needs and nuclear stakeholders' interests. Implementation aspects and prospects of Euratom FP6 are discussed bearing in mind the challenging additional objective. Very ambitious S/T actions have been undertaken supporting reactor safety not only by their technical achievements but also by their structuring effect in the European research picture. (authors)

  5. Light water reactor safety research project

    International Nuclear Information System (INIS)

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  6. Status of the fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    In 1967 and 1968 the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (''DeBeNe'') agreed to develop, in a joint program, breeder reactors to the point of commercial maturity. The following research organizations take part in this effort: Kernforschungszentrum Karlsruhe (KfK); INTERATOM, Bergisch Gladbach; ALKEM, Wolfgang near Hanau; SCK/CEN, Mol; Belgonucleaire, Brussels; ECN, Petten; TNO, Apeldoorn; NERATOOM, The Hague. The three German institutions mentioned above have been interrelated since 1977 by the Entwicklungsgemeinschaft (EG) Schneller Brueter. Between KfK, INTERATOM, and the French Commissariat a l'Energie Atomique contracts were concluded in 1977 about close cooperation in the Fast Breeder field, with association of the Belgian and Dutch partners. The results of research and development activities carried out by the DeBeNe partners in 1981 have been compiled in this report. The report begins with a short survey of the fast reactor plants, followed by an R and D summary. The bulk of the report gives more detailed information about those plants and about results reported by the Working Groups of the R and D Program Working Committee of the Fast Breeder Project. In an additional chapter a survey is given of international cooperation. (author)

  7. Summary of estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities

    International Nuclear Information System (INIS)

    A project is underway at Oak Ridge National Laboratory to assess the human health and environment effects associated with operation of Liquid Metal Fast Breeder Reactor fuel cycle. In this first phase of the work, emphasis was focused on routine radionuclide releases from reactor and reprocessing facilities. For this study, sites for fifty 1-GW(e) capacity reactors and three reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also, the results of these analyses are being used in the identification of research priorities. 13 refs., 2 figs., 3 tabs

  8. Estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities: a summary

    International Nuclear Information System (INIS)

    A project has been carried out at Oak Ridge National Laboratory to assess the human health and environmental effects associated with the operation of a liquid-metal fast breeder reactor fuel cycle. In this first phase of the work, emphasis was on routine radionuclide releases from reactor and reprocessing facilities. Sites for 51 1-GW(e) capacity reactors and 3 reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and the reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also the results of these analyses are being used in the identification of research priorities

  9. Recent progress in safety assessments of Japanese water cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Water Cooled Solid Breeder Test Blanket Module (WCSB TBM) is being designed by JAEA for the primary candidate TBM of Japan, and the safety evaluation of WCSB TBM has been performed. This reports presents summary of safety evaluation activities of the Japanese WCSB TBM, including nuclear analysis, source of RI, waste evaluation, occupational radiolysis exposure (ORE), failure mode effect analysis (FMEA) and postulated initiating event (PIE). For the purpose of basic evaluation of source terms on nuclear heating and radioactivity generation, two-dimensional nuclear analysis has been carried out. By the nuclear analysis, distributions of neutron flux, tritium breeding ratio (TBR), nuclear heat, decay heat and induced activity are calculated. Tritium production is calculated by the nuclear analysis by integrating distributions of TBR values, as about 0.2 g-T/FPD. With respect to the radioactive waste, the induced activity of the irradiated TBM is estimated. For the purpose of occupational radiolysis exposure (ORE), RI inventory is estimated. Tritium inventory in pebble bed of TBM is about 3 x 1012 Bq, and tritium in purge gas is about 3 x 1011 Bq. FMEA has been carried out to identify the PIEs that need safety evaluation. PIEs are summarized into three groups, i.e., heating, pressurization and release of RI. PIEs of local heating are converged without any special cares. With respect to heating of whole module, two PIEs are selected as the most severe events, i.e., loss of cooling of TBM during plasma operation and ingress of coolant into TBM during plasma operation. With respect to PIEs about pressurization, the PIEs of pressurization of the compartment nearby the pipes of cooling system are evaluated, because rupture of the pipes result pressurization of such compartments, i.e., box structure of TBM, purge gas loop, TRS, VV, port cell and TCWS vault. Box structure of TBM is designed to withstand the maximum pressure of the cooling system. At other compartments

  10. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    The proposed ATWS acceptance criteria for LWRs (as specified in NUREG-0460) are in principle, applicable to LMFBRs. For LMFBRs, the major criterion will be the assurance that the plant protection system (shutdown or scram) has a sufficiently high reliability (low failure rate) so that core disruptive accidents (as currently defined) will lie outside the design basis. For early plants, however, mitigating systems may also be required. Alternative accident scenarios for LMFBRs, which are initiated from the shutdown state or may lead to potential core disruptive accidents even following scram, need to be examined in greater detail. The proposed LWR-ATWS criteria do not appear to present any new or unforeseen design and/or safety questions for LMFBRs. They do, however, specify design goals for mitigating systems which may insure conformance with NRC policy. Preliminary recommendations are made for future research and evaluation

  11. Fusion reactor safety studies, FY 1977

    International Nuclear Information System (INIS)

    This report reviews the technical progress in the fusion reactor safety studies performed during FY 1977 in the Fusion Power Program at the Argonne National Laboratory. The subjects reported on include safety considerations of the vacuum vessel and first-wall design for the ANL/EPR, the thermal responses of a tokamak reactor first wall, the vacuum wall electrical resistive requirements in relationship to magnet safety, and a major effort is reported on considerations and experiments on air detritiation

  12. Deterioration of limestone aggregate mortars by liquid sodium in fast breeder reactor environment

    International Nuclear Information System (INIS)

    Highlights: • Limestone mortars were exposed to liquid sodium exposure at 550 °C. • Micro-analytical techniques were used to characterize the exposed specimens. • The performance of limestone mortar was greatly influenced by w/c. • The fundamental degradation mechanisms of limestone mortars were identified. - Abstract: Hot liquid sodium at 550 °C can interact with concrete in the scenario of an accidental spillage of sodium in liquid metal cooled fast breeder reactors. To protect the structural concrete from thermo-chemical degradation, a sacrificial layer of limestone aggregate concrete is provided over it. This study investigates the fundamental mechanisms of thermo-chemical interaction between the hot liquid sodium and limestone mortars at 550 °C for a duration of 30 min in open air. The investigation involves four different types of cement with variation of water-to-cement ratios (w/c) from 0.4 to 0.6. Comprehensive analysis of experimental results reveals that the degree of damage experienced by limestone mortars displayed an upward trend with increase in w/c ratios for a given type of cement. Performance of fly ash based Portland pozzolana cement was superior to other types of cements for a w/c of 0.55. The fundamental degradation mechanisms of limestone mortars during hot liquid sodium interactions include alterations in cement paste phase, formation of sodium compounds from the interaction between solid phases of cement paste and aggregate, modifications of interfacial transition zone (ITZ), decomposition of CaCO3, widening and etching of rhombohedral cleavages, and subsequent breaking through the weakest rhombohedral cleavage planes of calcite, staining, ferric oxidation in grain boundaries and disintegration of impurity minerals in limestone

  13. Development of electromagnetic pumps for natrium coolant of liquid metal fast breeder reactor (2)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Su, Soo Won; Kin, Hee Lyeong; Lee, Sang Doo; Seo, Joom Ho [Electrical Engineering and Science Research Institute, Seoul (Korea, Republic of)

    1994-07-15

    Present work on the development of annular linear induction pumps of externally-supported-duct type are to create domestic electromagnetic pumps by our own design and manufacturing technique and to secure the technological experience and data for the production of large scale electromagnetic pumps for natrium coolant loop system of liquid metal fast breeder reactor in the future. Two annular induction pumps, a small-sized one of 400 deg C and 60 l/min and a medium-sized one of 600 deg C and 800 l/min for their maximum operating temperatures and flowrates, respectively, are designed and fabricated. Conceptual and detailed designs for annular linear induction pumps with 60 l/min and 800 l/min flowrates, respectively, have been done by finding the optimum geometrical and operational parameters based on an equivalent-circuit analysis method. The measurements of the flowrates and pressures of the assembled pumps are done for confirming their characteristics and performance and comparing electrical input powers with those obtained from calculations. The cooling method developed in this study can be used in parallel with natural convection cooling without compressed air injection, and improves cooling efficiency and simplification of the pump structure. Experimental results measured by a free-fall indirect method and a EM flowmeter are and the design value of flowrate of each pump is confirmed by comparing measured one from indirect measurements. A center-return type pump for visualizing natrium pumping are also built with one pole pitch, eight outer core versions and six slots. Its natrium loop for pumping exhibition is assembled with instruments, heating equipment, leak sensing and pneumatic valve, and operated by a remote control. Magnetic flux distribution analysis is performed analytically and numerically for axial and radial directions in each case with or without end effects and consequently finds electromagnetic body force and pump efficiency.

  14. Fault tolerant distributed real time computer systems for I and C of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Architecture of distributed real time computer system (DRTCS) used in I and C of PFBR is explained. • Fault tolerant (hot standby) architecture, fault detection and switch over are detailed. • Scaled down model was used to study functional and performance requirements of DRTCS. • Quality of service parameters for scaled down model was critically studied. - Abstract: Prototype fast breeder reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Three-tier architecture is adopted for instrumentation and control (I and C) of PFBR wherein bottom tier consists of real time computer (RTC) systems, middle tier consists of process computers and top tier constitutes of display stations. These RTC systems are geographically distributed and networked together with process computers and display stations. Hot standby architecture comprising of dual redundant RTC systems with switch over logic system is deployed in order to achieve fault tolerance. Fault tolerant dual redundant network connectivity is provided in each RTC system and TCP/IP protocol is selected for network communication. In order to assess the performance of distributed RTC systems, scaled down model was developed with 9 representative systems and nearly 15% of I and C signals of PFBR were connected and monitored. Functional and performance testing were carried out for each RTC system and the fault tolerant characteristics were studied by creating various faults into the system and observed the performance. Various quality of service parameters like connection establishment delay, priority parameter, transit delay, throughput, residual error ratio, etc., are critically studied for the network

  15. Reactor safety in Eastern Europe. Proceedings

    International Nuclear Information System (INIS)

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP)

  16. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

  17. Activation calculation and environmental safety analysis for fusion experimental breeder (FEB)

    International Nuclear Information System (INIS)

    An activation calculation code FDKR and decay chain data library AFDCDLIB are used to calculate the radioactivity, decay heat, dose rate and biological hazard potential (BHP) form activation products, actinides and fission products in a Fusion Experiment Breeder (FEB). The code and library are introduced briefly, and calculation results and decay curves of related hazards after one year operation with 150 MW fusion power are given. The total radioactivity inventory, decay heat and BHP are 5.74 x 1020 Bq, 8.34 MW and 4.08 x 108 km3 of air, respectively, at shutdown. Results obtained show that the first wall of FEB can meet the nuclear waste disposal criteria for the NRC 10 CFR61 Class C after a few weeks from shutdown. The inventory of important actinides for the fuel reprocessing, such as 232U and 237Np were also calculated. It was shown that their concentrations do not excess the limit value of environmental safety required. (9 refs., 4 figs., 9 tabs.)

  18. Safety scenario for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    A scenario to ensure the safety of the Fusion Experimental Reactor (FER) is proposed. The safety features of a fusion reactor are given and their impacts on the safety design are shown. The requirements in the design of major components of FER to achieve safety and the safety evaluation process are described. The results of the evaluation showed that even in the event of the maximum credible accidents, the radiological consequence to the public can be held at an acceptable level. The applicability to FER of various aspects of the regulations for facilities treating large amounts of radioisotopes is discussed with a positive conclusion. (author). 11 refs, 1 fig

  19. Status of fast breeder reactor development in the United States of America - March 1986

    International Nuclear Information System (INIS)

    In a continuing situation of minimal electrical energy construction requirements to the turn of the decade, the U.S. LMR program has been modified to a modest level of technology development. Program focus is on innovative means to improve economics, provide inherent safety and environmental advantages and to stimulate involvement of utilities in ultimate demonstration ventures. The restructured technology program of reactor systems, core systems and fuels and materials continues at levels appropriate to provide the nucleus, and along with international collaboration, preserves the technology for utilization early in the next decade and beyond

  20. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used