WorldWideScience

Sample records for china experimental fast reactor

  1. China experimental fast reactor

    International Nuclear Information System (INIS)

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*1015 n/cm2/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  2. China experimental fast reactor; Le reacteur rapide experimental chinois

    Energy Technology Data Exchange (ETDEWEB)

    Tianmin, X. [Institut d' Ingenierie Nucleaire de Pekin (China); Cunren, L. [Centre d' Etude de Surete de Pekin (China)

    2007-07-15

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*10{sup 15} n/cm{sup 2}/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  3. Project management and its characteristic analysis for China experimental fast reactor

    International Nuclear Information System (INIS)

    As the first step of fast reactor development in China, China Experimental Fast Reactor (CEFR), one of the key projects of the National High Technology Research and Development Program (863 Program), will reach the first physical criticality in 2009 and will be connected to grid in 2010. The CEFR project includes R and D, design, construction, commissioning and operation, and its technology and management are very complex. This paper describes its position, objective, task, management, and analyzes its characteristics. (authors)

  4. Fast Reactor Development Strategy in China

    International Nuclear Information System (INIS)

    As one of the largest developing countries, China needs a reliable energy supplement. At the same time, China should improve the energy structure to decrease CO2 emissions. Nuclear and renewable energies are the main solutions to these issues. According to the research results, the nuclear capacity should increase to 400 GW(e) up to 2050. Fast reactors must be developed considering the limitation of uranium resources. In order to deploy fast reactor technology, the ‘experimental reactor, demonstration reactor and commercial reactor’ strategy has been suggested. China has finished the construction of the China Experimental Fast Reactor (CEFR) and gained necessary experience about fast reactors. The China Institute of Atomic Energy (CIAE) has begun to design the CFR-600, a 600 MW(e) demonstration fast reactor. This reactor will be put into operation before 2025. After that, a larger commercial reactor will be constructed. Besides fast reactors, all of other key sectors of fuel cycle will be developed at the same time such as reprocessing, fast reactor fuel, etc. There are two main tasks of fast reactors, one of which is to raise the utility ratio of uranium, and the other one is to transmute the long life waste of light water reactors. The fast reactor will be designed as a breeder and burner, respectively. (author)

  5. Analysis on vibration characteristics of the primary sodium pump of China experimental fast reactor

    International Nuclear Information System (INIS)

    The article introduces the rotational model analysis and the vibration test of the primary sodium pump of China Experimental Fast Reactor. Through the establishment of the rotation of the shaft system model, the critical speed has been analyzed. Combined with the pump bearings and motor bearings at double amplitude and RMS vibration velocity measurement experiment, the results show that the vibration characteristics of the primary sodium pump meet the requirements of the operational limits. (author)

  6. The verifying test of refueling system of the China experimental fast reactor

    International Nuclear Information System (INIS)

    The article introduce the verifying test of refueling system of China Experimental Fast Reactor. The purposes of the test is to check the performance of the equipment of refueling system, and to verify the requirement for the SCADA (Supervisory Control and Data Acquisition) system, and to verify the refueling SCADA system. For these purposes the test platform and device were built. For the first time in China, the simulated automated refueling was realized on the platform. This test has established the base for the test of refueling system on CEFR. (authors)

  7. Determination of hydrazine in third loops of China experimental fast reactor by spectrophotometry

    International Nuclear Information System (INIS)

    The method for the determination of hydrazine by Uv-vis spectrophotometer was proposed. The coloration conditions and instrument parameters were also optimized. In HCl, hydrazine formed a yellow azine with para-dimethyl aminobenzaldehyde ((CH3)2NC6H4CHO), and then determined by spectrophotometer. The complex's maximum absorption was exhibited at 458 nm. The coloration effect was excellent in conditions of 1% HCl, 10 mL para-dimethyl aminobenzaldehyde and 10 minutes' developing time. A good liner relationship was obtained in the range of 5∼200 μg/L, and the recovery was (101.1±1.9)%. This method was used in the third loop of China experimental fast reactor with satisfactory results. (authors)

  8. Seismic analysis of new fuel assembly loading machine for China experimental fast reactor

    International Nuclear Information System (INIS)

    New fuel assembly loading machine of China Experiment Fast Reactor is a kind of kinetic equipment with very complex structure. Many of its motional and driving components can not be simulated exactly by finite element model (FEM). A simplified FEM analysis method was introduced in the paper, and the main frame of the equipment was simulated by a simplified FEM model. Response spectrum analysis method was used to obtain the acceleration response of the main components of the equipment under seismic condition. Theoretical analysis method was used to calculate the stresses of the main connecting bolts, and these bolts were evaluated based the regulations of nuclear codes to ensure the structure integrity of the equipment. (authors)

  9. Fast Reactor and ADS development in China

    International Nuclear Information System (INIS)

    Conclusion: • The Fukushima accident influence China deeply. “The 12th five years plan and 2020 perspective goal of nuclear safety and radioactive pollution prevention” has been approved which means the nuclear may restart in the near future. • A demonstration fast reactor is under design. • More and more research works will be executed on CEFR

  10. Development of thermal-hydraulic steady-state analysis program for primary loop of China experimental fast reactor

    International Nuclear Information System (INIS)

    According to the characteristics of structure and steady-state for primary loop of China Experimental Fast Reactor (CEFR), a thermal-hydraulic steady-state analysis program was developed by using Fortran language. This paper focused on the development of a set of subroutine of physical properties of sodium and the sodium flow and heat transfer correlations for different operation conditions. And the difference among these correlations was compared. The calculation program was developed based on the steady model. At last, the thermal-hydraulic characteristics of steady-state of the primary loop of CEFR at full power were calculated. The calculation results are consistent with the design parameters and the correctness of the developed subroutines and steady- state calculation program was proved. (authors)

  11. Safety properties of China experimental fast reactor%中国实验快堆的安全特性

    Institute of Scientific and Technical Information of China (English)

    徐銤

    2011-01-01

    Sodium cooled fast reactor possesses some inherent safety properties, thanks to sodium perfect thermo-physical characteristics. In the same time sodium leakage inducing sodium fire or sodium-water reaction of industrial incidents, from sodium containing systems could not be excluded due to it is alkali metal. It is presented in the paper, that the safety of the China experimental fast reactor(CEFR)has meet the safety demands of Generation ]V due to the inherent safety characteristics have been realized, some passive safety systems, like passive decay heat removal system based on natural convection and circulation and active safety measures have been equipped. As for the large sized fast reactor with high breeding feature which induces positive sodium bubble effect, it is needed to develop passive shut-down systems to keep the safety targets of Generation IV.%钠冷快堆因钠具有好的热物理特性而具有固有安全性,同时也因钠是活泼的碱金属,也难免会有钠的泄漏、钠火和钠水反应等工业事故.本文介绍了中国实验快堆利用钠冷快堆的固有安全性,装设了单靠自然循环和自然对流的事故余热导出系统等多项非能动安全系统及完善的能动安全系统,其安全性达到了第Ⅳ代先进核能系统的安全要求.对于大型快堆,因其保证高的增殖而会有正的钠空泡效应,需要开发非能动停堆系统以保持第Ⅳ代安全目标.

  12. Fast Reactor Development Strategy in China

    International Nuclear Information System (INIS)

    China devotes herself to the peaceful use of nuclear to meet the growing energy demand. Proper amount of nuclear power plants could provide clean energy with low risk. • Fast reactor is a promising technology to ensure the sustainable development of nuclear energy, which can produce new fuel from depleted uranium and burn the long-life radioactive waste at the same time. It is expected that fast reactor will provide enough clean power to people for a long term in the future

  13. Fast reactor technology development in china status and prospects

    International Nuclear Information System (INIS)

    China has decided to speed-up the nuclear power development. It is programmed that the nuclear power capacity will reach 40 GWe in 2020 and envisaged 60 GWe and 240 GWe in 2030 and 2050 respectively. The basic strategy of PWR-FBR matched development with Fast reactor metal fuel closed cycle for a sustainable and quick increasing nuclear energy supply is adopted. Another strategy also decided is that the partitioning and transmutation of MA will be realized using fast burner and ADS. The fast reactor engineering development will be divided into three steps: China Experimental Fast Reactor (CEFR 65 MWt/20 MWe), 3China Prototype/Demonstration Fast Reactor (CPFR/CDFR ≥1 500 MWt/600 MWe) and China Demonstration Fast Breeder Reactor (CDFBR 1 000-1 500 MWe). The CEFR is under installation and pre-operation testing with it's first criticality planned in 2009. The design study of CPFR is just started in 2006. Recently a discussion for the second step is under way to faster the fast reactor development by a larger than 600 MWe CPFR and as a role of CDFR. (authors)

  14. Hydraulic Experiment for Simulative Assemblies of Blanket Assembly and Np Transmutation Assembly of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    CHENG; Dao-xi; QI; Xiao-guang; ZHAI; Wei-ming; YANG; Bing; ZHOU; Ping

    2013-01-01

    The out-of reactor hydraulic experiment of fast reactor assembly is one of the important experiments in the process of the development of the fast reactor assembly.In this experiment,the size of the throttling element in the foot of the assembly is decided which is fit for the flow division in the reactor and the

  15. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    The paper outlines the recent development status of nuclear power plants in China and introduces the main design characteristics and nuclear safety features of the Chinese Experimental Fast Reactor (CEFR). During the review of the Preliminary Safety Analysis Report some important subjects have been proposed by the China National Nuclear Safety Administration (NNSA). More detailed research for the answer has been done. The main analysis results for (1) Reactor Shut-down System, (2) Decay Heat Removal System and (3) Fuel Subassembly Blockage as three examples are given in this paper. The CEFR is still in the detail design stage. Its site is almost ready for the construction of the main building. It is planned to have the first pouring of concrete in June, 1999, but it depends on the license issued by the NNSA. (author)

  16. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author)

  17. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  18. Fast reactor development for sustainable nuclear energy supply in China

    International Nuclear Information System (INIS)

    China needs a very huge energy supply for national economy development and living standard improvement of 1.3 billion population. The nuclear energy is a new member of the energy supply family in China. A satisfied operation records of all 11 units of NPPs, especially with the total average load factor 85.8% of all NPPs in 67 reactor-years since commercial operation of each unit encourage the public to believe that the nuclear power is a safe, reliable, economically-acceptable, CO2 avoidable one and could be used in large scale. The government has decided in 2006 to accelerate the nuclear power development with the target of 40GWe in operation and 18GWe in construction in 2020.Right now 13 units with total capacity 13.05GWe are under construction and other 11 units of total capacity 12.05GWe have been approved by the government and the preparation for construction is underway. For the sustainable supply of nuclear energy, as the principle strategy, PWR-FBR matched with closed nuclear fuel cycle has been decided by the government for a long time. Three FBR development strategy targets suggested as following. (1) to realize FBR commercial utilization in small batch in 2030; (2) to increase nuclear capacity to 240GWe, sharing about 16%, mainly by FBRs in 2050, and (3) to replace coal fired plants by nuclear power in large scale, in the period about 2050-2100. For that, the suggested FBR development strategy is shown in Table 1. China Experimental Fast Reactor (CEFR) with the power 65MWt is a pool type sodium-cooled fast reactor. Pre-conceptual design started in 1990 with the first pot of concrete in 2000, and architecture engineering launched in 2001. Now it is under commissioning tests stage. The experiences of design fabrication and construction for this type of reactor have been gained. The possibility of large striding from 20MWe CEFR to 800MWe China Demonstration Fast Reactor (CDFR) has been studied. The favorableness is estimated mainly as following. (1) The

  19. The status and prospects of fast reactor technology development in china

    International Nuclear Information System (INIS)

    China has decided to speed-up the nuclear power development. It is programmed that the nuclear power capacity will reach 40 GWe in 2020 and envisaged 60 GWe and 240 GWe in 2030 and 2050 respectively. The basic strategy of PWR-FBR matched development with fast reactor metal fuel closed cycle for a sustainable and quick increasing nuclear energy supply is adopted. Another strategy also decided is that the partitioning and trans-mutation of MA will be realized using fast burner and ADS. The fast reactor engineering development will be di-vided into three steps: China Experimental fast Reactor (CEFR 65 MWt/20 MWe), China Prototype/Demonstration Fast Reactor (CPFR/CDFR≥1 500 MWt/600 MWe) and China Demonstration Fast Breeder Reactor(CDFBR100-1500 MWe). The CEFR is under installation and pre-operation testing with its first criticality panned in 2009. The design study of CPFR is just started in 2006. Recently a discussion for the second step is under way to faster the fast reactor development by a larger than 600 MWe CPFR and as a role of CDFR. (authors)

  20. China Experimental Fast Reactor(CEFR)——Criterion of Criticality for Reactor With External Neutron Source

    Institute of Scientific and Technical Information of China (English)

    ZHAOYu-sen

    2003-01-01

    There is a neutron source with 109 s-1 neutrons in core of CEFR during start up test and operation of CEFR. For judging the criticality of reactor with external neutron source and near criticality, it is important that the neutron level changes in core with time must be understood after introducing positive reactivity to core with external neutron source.

  1. Fast reactor development strategy targets study in China

    International Nuclear Information System (INIS)

    China is a big developing Country who needs a huge energy resources and a rapid growing rate. Considering energy resources limited and environment issues it is sure that the nuclear energy will be becoming one of the main energy resources. The Government has decided to develop the nuclear power capacity to 40 GW in 2020. It is envisaged that it will reach to 240 GW in 2050. It is stimulate us to consider conscientiously the development of the fast breeder reactor's and related closed nuclear fuel cycle by the limitation of Uranium resources and uncertainties of international Uranium market. Followings are the proposed strategic targets of fast reactor development in China. (1) To realize the operation of commercial fast breeder reactors with an unit size of 800-900 MWe and one site-multi reactors in 2030. (2) To develop the nuclear power capacity to 240 GW in 2050. (3) To replace step by step the fossil fuel utilization in large scale by nuclear energy beyond 2050. (authors)

  2. 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

  3. Analysis of sodium/argon flow in anti-siphon equipment's curved pipeline of China Experimental Fast Reactor

    International Nuclear Information System (INIS)

    According to the study of CEFR anti-siphon equipment structure and its two-phase flow phenomenon, we can deeply understand its passive function which could be used to reduce the leakage of sodium. Compared to the experimental data,a numerical model can correctly describe this phenomenon of leakage processes. (authors)

  4. Fast Reactor Development for a Sustainable Nuclear Energy Supply in China

    International Nuclear Information System (INIS)

    Nuclear energy is a new member of the energy supply family in China. Satisfactory operating records of all 11 nuclear power plants in China encourage its stepwise and large scale use and the PWR-FBR route matched with a closed nuclear fuel cycle forms a basic strategy. The sufficient utilization of nuclear resources and the treatment of highly radioactive waste by transmutation in fast reactors are the key issues for a sustainable development of nuclear energy. As the first step in FBR engineering development, the 65 MW(th) China Experimental Fast Reactor is approaching startup, the conceptual design of the 600-900 MW(e) China Demonstration Fast Reactor (CDFR) has been started and the 1000-1500 MW(e) China Demonstration Fast Breeder Reactor is under consideration. Three FBR development strategy targets are as follows: (1) To start realizing CDFR type commercial utilization in small batches by 2030; (2) To increase nuclear capacity to 240-250 GW(e), representing about 16%, mainly through FBRs by 2050; (3) To replace coal fired plants by nuclear power on a large scale in the period 2050-2100. (author)

  5. The SCARABEE experimental fast reactor safety programme already completed

    International Nuclear Information System (INIS)

    The SCARABEE in-pile experimental programme comprised a series of tests on unirradiated fuel pins, either single or in seven-pin clusters. The main objective was to obtain information on the mode and consequences of fast reactor fuel pin failure in conditions representative of loss of cooling in a LMFBR. The application of such programmes in full scale reactors leads to the great importance of the interpretation of experimental observations. The interpretation of that programme was carried out jointly by CEA, KFK and UKAEA; this international collaboration led to a sharper focusing on essential features to be modelled in experiments and computer codes and to a valuable convergence of views

  6. Design Features and Operating Experience of Experimental Fast Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The IAEA has begun an initiative to help coordinate Member State efforts in the field of fast neutron nuclear reactors. This initiative is primarily targeted at the preservation of knowledge in the areas of design, construction and operation, for both experimental and power fast reactors. The ultimate goal of this activity is to establish a comprehensive, international inventory of fast reactor data and knowledge, which will be an essential resource for the future development and deployment of fast reactor technology. In this project, carried out within the framework of the

  7. Seismic appraisal test of control rod drive mechanism of China experiment fast reactor

    International Nuclear Information System (INIS)

    The structure of the control rod drive mechanism in pool type sodium-cooled fast reactor is the characterized by long, thin, and geometric nonlinearity, and the seismic load is multiple activation. The anti-seismic evaluation is always paid great attention by the countries developing the technology worldwide. This article introduces the seismic appraisal test of the control rod drive mechanism of China Experimental Fast Reactor (CEFR) performed on a seismic platform which is vertical shaft style and multiple activation. The result of the test shows the structural integrity and the function of the control rod drive mechanism could meet the design requirements of the earthquake intensity. (authors)

  8. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  9. Project and characteristics of a 5MW experimental fast reactor

    International Nuclear Information System (INIS)

    Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 1015 n/cm2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author)

  10. 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.)

  11. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    Since the last annual meeting of the IWGFR, Chinese nuclear power plants, especially Qing Shan 300 MWe, have maintained rather good operation records. The Chinese Experimental Fast Reactor (CEFR) with the power 65 MWt, 25MWe is still in the preliminary design stage. The results based on its conceptual design are given in this paper. Few R and D results concerning materials, instrumentation, neutronics study and sodium loops, have been obtained last year due to lack of budgets. As an important progress for the CEFR project, the State Planning Commission has approved this project at the end of the year 1995. It means that after the approval of its feasibility study reports, the CEFR will be becoming a National Annual Project. (author)

  12. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J. [Phenix Plant (France)

    2007-07-01

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  13. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    International Nuclear Information System (INIS)

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  14. Transient simulation code development of primary coolant system of Chinese Experimental Fast Reactor

    International Nuclear Information System (INIS)

    Highlights: ► A transient analysis code is developed for Chinese Experimental Fast Reactor. ► A set of subroutines for friction and heat transfer correlations were compiled. ► The calculation speed of this code is fast enough for real-time simulation. - Abstract: Chinese Experimental Fast Reactor (CEFR) is a 25 MWe sodium cooled, pool type reactor, which was built at the China Institute of Atomic Energy in Beijing as the forerunner to the first-stage of Chinese fast reactor development plans. In order to understand the response of the Primary Coolant System (PCS) to various transients and train the operators a dynamic model using basic energy and momentum equations was developed with some assumptions. Heat transfer models for reactor core and intermediate heat exchanger were also included. Subroutines were developed to calculate the thermal properties, friction coefficients and heat transfer coefficients of liquid sodium. Gear’s method was applied to solve the dynamic model. A transient analysis code named THPCS (Thermal–Hydraulic code of PCS) was developed and is independent of the design and safety analysis codes. Three typical events, such as loss of one primary pump, unprotected transient overpower and accidental closure of primary pump check valves were chosen and investigated. The prediction results of the code agree well with those of the final safety analysis report of CEFR. A fourth postulated accident, station blackout without scram and loss of all heat sink, was also analyzed to show the ability of the code, which is more serious than the former. The transient simulation code developed in this paper will be useful for the safety operation of CEFR

  15. The status of fast reactor technology development and accelerator driven subcritical system researches in China

    International Nuclear Information System (INIS)

    Since last May in mainland China there are two nuclear power plants with total capacity of 2.1 GWe in operation and four NPPs in construction. It is envisaged that the total nuclear power capacity will be about 8.5 GWe in the year 2005. Recently the Government is considering four other new NPPs with a total capacity of about 4 GWe and starting their construction during 'tenth five years Plan' (2001-2005). The three new nuclear systems, FBR, ADS and Hybrid, have started to be developed with a rather moderate project and are all still in the early stage. For fast reactor engineering development, the China Experimental Fast Reactor (CEFR) of 65 MWt is the first step. After some additional accidents analysis, especially sodium spray fire accident analysis, the reactor building construction will be continued. The main components including of the reactor block, primary and secondary circuits, fuel handling system have been ordered. It is foreseen to have CEFR reaching first criticality at the end of 2005. The second step 300 MWe Modular Fast Reactor (MPFR) is under consideration, which will be a prototype for large size fast reactor. Based on the size of MPFR, the role of MA transmutation has been evaluated. For the Accelerator Driven Subcritical System (ADS), we are making great efforts to accomplish the research tasks worked out in the first phase program (1998-2002) with emphasis on the system optimization, reactor physics and technology, accelerator physics and technology and nuclear and material data base, and are enthusiastically preparing to step to the second phase program which is marked by ADS concept verification study (2000-2007). As to the Fusion-Fission Hybrid System, in near-term the emphasis will be put on the experiments on two big testing facilities HL-1M and HT-7 on one hand, and on the other hand, we will determine the targets of medium-term and long-term development for Hybrid system and work out relative development program

  16. 中国实验快堆全堆芯流量分配计算与试验%Calculation and Test of Core Flow Rate Distribution of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘一哲; 薛秀丽; 许义军; 冯预恒; 侯志峰

    2012-01-01

    Based on the core and primary circuit design of China Experimental Fast Reactor(CEFR), a multiple-channel thermal-hydraulic analysis code DAEMON was developed to calculate the core flow rate distribution and unsymmetric coefficient in different conditions. In the commissioning stage, a series of full-scale tests for reactor core were performed in CEFR with a permanent-magnet sodium flow meter. The numerical results of code DAEMON showed a good agreement with test data. The core hydraulic design was also validated with a view to the requirements of design criteria, commissioning and operation specifications.%针对中国实验快堆(CEFR)堆芯和一回路的设计特点,开发水力特性计算程序DAEMON,完成不同工况下的全堆芯流量分配计算,给出流量分配不均匀性等参数.在反应堆调试阶段,进行全堆芯流量分配试验.结果表明,程序计算值与试验值符合较好.在此基础上,验证了CEFR堆芯的流体力学设计,并为反应堆调试和运行提供了基础数据.

  17. 中国实验快堆全厂断电事故多维度热工耦合计算%Multi-dimension Coupled Simulation Method of Thermalhydralic Behavior in China Experimental Fast Reactor Under Blackout Accident

    Institute of Scientific and Technical Information of China (English)

    乔雪冬; 胡文军; 冯预恒; 张春明; 孙微; 赵守智

    2012-01-01

    多维度耦合方法是将传统的一维反应堆热工流体力学程序与三维流体动力学分析软件通过一定的耦合方法结合起来,实现反应堆局部复杂流体现象分析与系统计算的耦合方法.本工作根据中国实验快堆设计和运行经验,开发了基于Rubin和Fluent的耦合程序框架,完成了中国实验快堆全厂断电工况的计算和验证.计算结果表明,耦合方法对全场断电事故的计算结果合理可靠,是对一维系统程序分析方法的有益补充.%Multi-dimension coupled simulation is a method which combines the analysis of complex hydromechanical phenomenon in reactor with system calculation by the method of coupling traditional one-dimensional thermo hydrodynamic program with CFD software. The coupling frame was developed based on Rubin and Fluent codes. By the test calculation under the station blackout accident of China Experimental Fast Reactor (CEFR) , multi-dimension coupled simulation is proved reasonable and gives a efficient supplement to system calculation method.

  18. 中国实验快堆典型钠阀温度分布研究%Study on temperature distribution in bellows seal sodium valve assembly of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    李生; 张东辉; 刘云焰

    2012-01-01

    The typical bellows seal sodium valves, as the important equipment in the sodium systems of China Experimental Fast Reactor, have a significant effect on the safety of fast reactors. The typical valves caused some problems in the test stage. The paper is to get the results of temperature distribution in the sodium valve assembly on stable heat transfer condition, using the Computational Fluid Dynamics (CFD) code. The paper also analyses the results computed under the condition of fixed thickness and the different height of thermal insulation materials. It gets a good conclusion through the comparison of measured and simulated results and the numerical simulation result is logical and meaningful.%中国实验快堆典型钠阀作为系统重要的涉钠设备,直接影响着反应堆系统的安全运行.中国实验快堆工程在调试和运行阶段面临着钠阀门带来的一系列问题.本文应用CFD软件计算了两种运行工况下典型钠阀稳态温度场分布,分析了保温层厚度一定、高度不同的情况下,钠阀门的温度场分布结果,并与实验结果进行了对比,证明结果是合理的有意义的.

  19. 中国实验快堆主蒸汽系统优化设计及分析研究%Optimization design and analysis on main steam system of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    纪西胜; 吴强; 牛敬娟

    2012-01-01

    The function of the main steam system of China Experimental Fast Reactor (CEFR) is to transfer the steam from SG to the turbine to generate power and to discharge the steam from reactor in case of accidental condition. However, the reactor automatically shut down many times due to improper operation of valves, affecting the stability of the system and increasing operation cost. In order to optimize the steam system process flow, this paper introduced the bypass and worked out the design parameters, and finally gave qualitative analysis of the calculation result in special condition.%中国实验快堆三回路主蒸汽系统主要功能是将蒸汽发生器产生的蒸汽送至汽轮发电机组,辅助功能是在事故工况下排出反应堆产生的热量.调试期间多次因主蒸汽系统阀门手动操作而引起停堆,影响了系统的稳定性,增加了运行成本.本文对主蒸汽系统进行了优化设计、增加旁路管道,并对此条件下的过热器反暖操作和特殊工况下压力损失计算结果进行定性分析,确定了设计参数,优化了主蒸汽系统的工艺流程.

  20. A Study of Reactor Neutrino Monitoring at Experimental Fast Reactor JOYO

    CERN Document Server

    Furuta, H; Hara, T; Haruna, T; Ishihara, N; Ishitsuka, M; Ito, C; Katsumata, M; Kawasaki, T; Konno, T; Kuze, M; Maeda, J; Matsubara, T; Miyata, H; Nagasaka, Y; Nitta, K; Sakamoto, Y; Suekane, F; Sumiyoshi, T; Tabata, H; Takamatsu, M; Tamura, N

    2011-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3m from the JOYO reactor core of 140MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11\\pm1.24(stat.)\\pm0.46(syst.)events/day. Although the statistical significance of the measurement was not enough, the background in such a compact detector at the ground level was studied in detail and MC simulation was found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  1. Hydrodynamics Analysis of Inlet in Anti-siphon Equipment of China Experimental Fast Reactor%中国实验快堆虹吸破坏装置取钠口结构流体动力学分析

    Institute of Scientific and Technical Information of China (English)

    彭燕; 张东辉; 丁振鑫

    2011-01-01

    中国实验快堆在一回路钠净化系统中设置虹吸破坏装置,以非能动方式减少该系统发生堆外管道破裂事故的液态钠泄漏量.本文对该装置中取钠口结构的发泡动力效应进行研究,从流体动力学分析角度证实该装置改进结构取钠口的泄压能力和非能动减少液态钠泄漏量的能力比原结构取钠口的好.%The anti-siphon equipment is set in the primary sodium purification system of China Experimental Fast Reactor, which is used to passively reduce the sodium leakage when this system suffers any accident of the system pipeline breaks. From the point of view of bubble growing theory, the hydrodynamics analysis for the inlet of this equipment was performed. It is certified that the ability of the improved inlet, which includes depressurization and reducing the leak quantity of liquid, is better than the original one.

  2. Linear and nonlinear stability analysis, associated to experimental fast reactors. Part 2

    International Nuclear Information System (INIS)

    The nonlinear effects in fast reactors kinetics and their stability are studied. The Lyapunov criteria and the Lurie-Letov functions for nonlinear systems were established and simulated. Small oscillations were studied by a Fourier analysis to clarify particular aspects of feedback and load functions in fast reactor at zero power, or/and in normal power level. The results were in agreement with the experimental data existing in the literature. (E.G.)

  3. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko;

    2009-01-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic...

  4. Development of Thermal-Hydraulic Steady-State Analysis Program for Primary Loop of China Experimental Fast Reactor%中国实验快堆一回路热工水力稳态计算程序开发

    Institute of Scientific and Technical Information of China (English)

    饶彧先; 崔满满; 郭赟

    2012-01-01

    针对中国实验快堆(CEFR)的具体结构和稳态运行特点,利用Fortran语言开发了CEFR一回路热工水力稳态计算程序.重点开发了有关钠的多种物性的子程序、适应不同工况的钠的流动与换热计算子程序,并对关系式进行了对比分析,最后建立了稳态计算模型并开发了程序.在此基础上,对CEFR的一回路系统在满功率下的稳态热工水力特性进行了计算分析,所获得的结果同设计参数吻合,证明了所开发的子程序及稳态程序的正确性.%According to the characteristics of structure and steady-state for primary loop of China Experimental Fast Reactor (CEFR), a thermal-hydraulic steady-state analysis program was developed by using Fortran language. This paper focused on the development of a set of subroutine of physical properties of sodium and the sodium flow and heat transfer correlations for different operation conditions. And the difference among these correlations was compared. The calculation program was developed based on the steady model. At last, the thermal-hydraulic characteristics of steady-state of the primary loop of CEFR at full power were calculated. The calculation results are consistent with the design parameters and the correctness of the developed subroutines and steady-state calculation program was proved.

  5. Sodium Fi re Probabilistic Safety Assessment of Primary Cold Trap Room for China Experimental Fast Reactor%中国实验快堆一回路冷阱工艺间钠火概率安全评价

    Institute of Scientific and Technical Information of China (English)

    宋维; 胡文军; 钱鸿涛; 付陟玮; 左嘉旭

    2015-01-01

    本文运用事件树方法对中国实验快堆一回路冷阱工艺间发生钠火后的事故场景进行演绎分析,运用故障树方法对钠火相关系统进行可靠性建模。在此基础上计算得到各钠火事故序列的条件发生概率。结果表明:在获得的25个典型钠火事故序列中,19个序列的条件发生概率较低;在发生概率相对较高的6个序列中,4个序列的后果轻微,其余两个序列代表的钠火场景存在一定不确定性,需要在今后的钠火危险性评价中进一步具体研究。%The sodium fire scenarios after the fire was ignited in primary cold trap room of China Experimental Fast Reactor were deduced using event tree method .The systems related to the accident were modeled using fault tree method .Thereby ,the conditional occurrence probabilities of all sodium fire sequences were calculated .The results show that 25 typical sodium fire accident sequences are obtained in total ,and 19 sequences have lower probabilities of occurrence .Among the 6 sequences with relatively higher probabilities ,4 sequences cause minor consequences , and the remaining 2 sequences require a detailed hazard evaluation in the next work because of the uncertainty .

  6. Research on Startup Condition of Steam Generator in China Experimental Fast Reactor%中国实验快堆蒸汽发生器启动工况研究

    Institute of Scientific and Technical Information of China (English)

    吴强; 纪西胜; 牛敬娟; 张焕旗; 谢海昕

    2015-01-01

    The steam generator in China Experimental Fast Reactor (CEFR) is direct flow type ,and the steam generator startup of CEFR is quite different and more compli‐cated than that of PWR .In this paper ,the startup condition of steam generator was studied ,and the operation parameters were compared with theoretical design parame‐ters .The results show that the calculated theoretical value of operation parameters is basically coincident with test value .The optimized startup program of steam generator for guiding operator was also introduced .The optimized procedure to start up the steam generator is proposed for guiding operators ,and it effectively ensures the power opera‐tion test of CEFR .%中国实验快堆蒸汽发生器为直流式,启动方式与压水堆核电厂的有较大区别,启动过程较为复杂。本文对中国实验快堆蒸汽发生器启动工况进行了研究,并将运行参数与理论设计参数进行了比较。结果表明,运行参数理论计算值与试验值基本吻合。提出了蒸汽发生器启动运行的优化方案,以指导运行人员操作,有效地保障了中国实验快堆功率运行试验的开展。

  7. In-pipe experimental needs for resolution of principal reactor safety issues in commercialization of fast reactors

    International Nuclear Information System (INIS)

    This paper describes major research subjects and approaches on reactor safety for commercialization of fast reactors including recriticality issue in core disruptive accident sequences. To achieve the research objective for these major subjects, a new in-pile safety experimental program named SERAPH (safety engineering reactor for accident phenomenology) is proposed, and the conditions of the proposed tests and the major requirements for the facility are formulated. 13 refs., 5 figs., 1 tab

  8. Out-of-pile experimental stand for research of fast reactors safety questions

    International Nuclear Information System (INIS)

    In the given paper there is given the description of experimental facility 'EAGLE', which is meant for research of safety problems of fast reactors one of which is excluding of re-criticality in case of severe accident with core melting. There are demonstrated concepts and volume of planned tests and also the results of conducted experiments. (author)

  9. Fast reactor database. 2006 update

    International Nuclear Information System (INIS)

    Liquid metal cooled fast reactors (LMFRs) have been under development for about 50 years. Ten experimental fast reactors and six prototype and commercial size fast reactor plants have been constructed and operated. In many cases, the overall experience with LMFRs has been rather good, with the reactors themselves and also the various components showing remarkable performances, well in accordance with the design expectations. The fast reactor system has also been shown to have very attractive safety characteristics, resulting to a large extent from the fact that the fast reactor is a low pressure system with large thermal inertia and negative power and temperature coefficients. In addition to the LMFRs that have been constructed and operated, more than ten advanced LMFR projects have been developed, and the latest designs are now close to achieving economic competitivity with other reactor types. In the current world economic climate, the introduction of a new nuclear energy system based on the LMFR may not be considered by utilities as a near future option when compared to other potential power plants. However, there is a strong agreement between experts in the nuclear energy field that, for sustainability reasons, long term development of nuclear power as a part of the world's future energy mix will require the fast reactor technology, and that, given the decline in fast reactor development projects, data retrieval and knowledge preservation efforts in this area are of particular importance. This publication contains detailed design data and main operational data on experimental, prototype, demonstration, and commercial size LMFRs. Each LMFR plant is characterized by about 500 parameters: physics, thermohydraulics, thermomechanics, by design and technical data, and by relevant sketches. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors with complete technical information of a total of 37 LMFR

  10. Operational experience and upgrading program of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Twenty years of successful operations at the experimental fast reactor JOYO provide a wealth of experience covering core management, chemical analysis of sodium and cover gas for impurity control, natural convection tests, upgrade of fuel failure detection system, corrosion product measurement, development of operation and maintenance support system, and replacement of major components in the cooling systems. Some of the data obtained is stored in a database to preserve the related knowledge. This experience and accumulated data will be useful for the design of future fast reactors. (author)

  11. Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

    International Nuclear Information System (INIS)

    The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D2O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

  12. Restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    The experimental fast reactor Joyo is the first sodium cooled fast reactor in Japan. Joyo attained initial criticality as a breeder core in April 1977 and has operated as a high performance irradiation test bed since 2003. The 15th periodic inspection of Joyo commenced in May 2007 with the Fuel Handling Machine (FHM) being set up on the Rotating Plug (R/P) for refueling in June. When the R/P was taken down, measuring the load of the Hold-Down Shaft (HDS) revealed an abnormal decrease above the in-vessel storage rack (IVS). The HDS is a cylindrical FMH device that holds down the 6 surrounding subassemblies (S/As) which are adjacent to a withdrawn S/A. In order to investigate the cause of this, an in-vessel observation was conducted using a radiation-resistant fiber scope (RRF). As a result of the observations, it was discovered that the top of the irradiation test S/A 'MARICO-2' (the material testing rig with temperature control) had bent onto the IVS as an obstacle, and had damaged the Upper Core Structure (UCS). During the investigation of this incident, the in-vessel observations using RRF etc. took place at (1) the top of the S/As and the IVS for foreign material, (2) the bottom face of the UCS for damage under the condition with the level of sodium at -50 mm below the top of the S/As. In-vessel observation techniques for a Sodium cooled Fast Reactor (SFR) are important in confirming its safety and integrity. Since an in-vessel observation for an SFR has to be conducted under severe conditions that include high temperatures (∼ 200 deg-C) and high radiation doses (∼ 400 Gy/h), and the primary sodium coolant has to be retained in the Reactor Vessel (R/V) to remove the decay heat, an in-vessel observation equipment has to be designed to not only tolerate the severe conditions but also be capable of being inserted into the sealed R/V through the fixed holes built in to the R/P and gain access to the observation areas. The in-vessel observations were successfully

  13. 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

  14. 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)

  15. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  16. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    International Nuclear Information System (INIS)

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 ± 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF

  17. A new concept of laser fusion experimental reactor with fast ignition target

    International Nuclear Information System (INIS)

    Full text: We have analyzed the design windows of laser fusion power plants based on fast ignition targets, and examined feasibility of a small-sized laser fusion experimental reactor suitable for developing their power plants. Target gain curves are evaluated for power plants, which have 100∼200MJ fusion yields with 600kJ∼1MJ lasers, and for an experimental reactor (LFER), which has a 10MJ fusion yield with a 200kJ laser, 100kJ for implosion and 100kJ for heating. The pulse heat loads on the chamber wall of LFER are estimated at 2.5J/cm2 for a 2.5-m-radius solid wall chamber, and 16J/cm2 for a 1-m-radius liquid wall chamber. The fast ignition LFER can make its fusion output one order smaller than that of the central ignition, thus we can use a rather small solid wall chamber for the first stage of the LFER. We can also expect to decrease laser cost drastically, although for a heating laser we must develop the long life final optics. Through a fast ignition LFER, we showed a possibility to demonstrate net electric generation in a reasonably short time. (author)

  18. A new concept of Laser Fusion Experimental Reactor with fast ignition target

    International Nuclear Information System (INIS)

    We have analyzed the design windows for laser fusion power plants based on fast ignition concepts, and examined the feasibility of a small-sized laser fusion experimental reactor suitable for developing their power plants. Target gain curves are assessed for power plants, having 90∼200 MJ fusion yields with 600 kJ∼1MJ lasers, and for an experimental reactor (LFER), having a 10 MJ fusion yield with a 200 kJ laser, i.e., 100 kJ for implosion and 100 kJ for heating. The pulse heat loads on the chamber wall of LFER are estimated as 2.5 J/cm2 for a 2.5-m-radius solid wall chamber, and 16 J/cm2 for a 1-m-radius liquid wall chamber. The fast ignition LFER can produce its fusion output approximately one order of magnitude smaller than that of the central ignition, so that we can use a rather small solid wall chamber for the first stage of the LFER operation. We can also expect to decrease laser cost drastically, although for the heating laser we must develop a long life final optics system. With the fast ignition LFER, we showed a possibility to demonstrate net electric generation in a reasonably short time. (author)

  19. Development of a transient thermal-hydraulic code for analysis of China Demonstration Fast Reactor

    International Nuclear Information System (INIS)

    Highlights: ► A transient thermal-hydraulic code is developed for analysis of CDFR. ► The code to code validation shows good accuracy and reliability of the code. ► Multiple-channel model is applied to depict the core. ► Compressible homogenous flow model is used for the two-phase flow of sodium. - Abstract: The transient thermal-hydraulic code THACOS is under development for analysis of China Demonstration Fast Reactor. Applying modular technology, the code contains the core module, the pump module, the sodium pool module and the heat exchanger module and each module could operate separately. It can provide one-dimensional thermal-hydraulic simulation for the primary sodium coolant loop. The point reactor kinetics equations with six-group delayed neutrons have been applied to calculate the core power considering reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of the core. Multiple-channel model is applied to depict the core. Compressible homogenous flow model is used for the two-phase flow of sodium. The calculated results show that sodium boiling will occur quickly under the ULOF accident without any shutdown rods insertion. While, with the insertion of three hydraulically suspended shutdown rods, the core could be shut down safely and boiling will not occur in a short period of time. Obviously, the passive hydraulically suspended shutdown rods could keep the core safe under ULOF accidents

  20. Renovation of new fuel transfer machine in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    In the higher performance plan (MK-III plan) of the experimental Fast Reactor JOYO, fuel handling system has been renovated to remote control system to reduce refueling time. As a part of this plan, new fuel transfer machine which is used to receive and transport new fuel, has been renovated completely to remote an automatic control system with no local operation and no local watching by Kawasaki Heavy Industries, Ltd. In this paper, the design and fabrication of this system are described. (author)

  1. Experimental software design of neutron texture diffractometer at China advanced research reactor

    International Nuclear Information System (INIS)

    The experimental software of the neutron texture diffractometer at China Advanced Research Reactor (CARR) was designed. Based on the principle of texture measurement by neutron diffraction and the motion control and data acquisition system of the diffractometer, the functions needed for texture measurement were proposed. Then the flow charts of these functions were described in detail and realized by Python language in Linux system. The experimental software for CARR neutron texture diffractometer has been successfully accomplished. (authors)

  2. Student Training Course Using the Experimental Fast Reactor JOYO and Related Facilities

    International Nuclear Information System (INIS)

    University level training courses have been initiated and implemented using the Experimental Fast Reactor Joyo and related facilities of the Japan Atomic Energy Agency (JAEA). These courses offer nuclear facility on-site education and experience in conjunction with a highly experienced engineering staff. University Nuclear Engineering Department faculty members have strongly supported and collaborated in the development of this program. The program covers reactor core physics analysis plus experiments using full-scope training simulator and performing neutron dosimetry, isotopic analysis of noble gases, chemical analysis of sodium, etc. This program is also anticipated to promote the human resource development in the younger generation for the nuclear industry, and to strengthen the relation between JAEA and University research programs. (author)

  3. Characterization of irradiation fields for fuel and material irradiation in the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    The Joyo MK-III core is a worldwide fast neutron irradiation field not only for FBR development but also for use in other fields such as light water reactor (LWR) and fusion reactor studies, and in the non-nuclear industry. The characterization of these neutron and gamma ray fields is most important to utilize for irradiation tests. This paper describes the details of distributions of neutron flux, reaction rate and gamma heating in the MK-III core. The calculation accuracy of the core management codes HESTIA, TORT and MCNP, was also evaluated by the measured data. The calculated results in neutron calculation agreed well with the measured one. The calculation method was validated and correction factors were identified. In case of gamma heating evaluation, the calculated result is underestimated with respect to the experimental value especially in the upper and lower SS reflector region. Further investigations in gamma heating evaluation are needed. (author)

  4. Experimental modelling and numerical analysis of a molten salt fast reactor

    International Nuclear Information System (INIS)

    In this paper experimental and numerical investigation of the MSFR (Molten Salt Fast Reactor) concept will be presented. This homogeneous, single region liquid fuelled fast reactor concept uses fluoride-based molten salts with fissile uranium and thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. Molten salt reactors with liquid fuel have a unique safety related property that needs clear understanding. In the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR concept has a single region homogeneous core without internal structures, it is a difficult task to ensure stable flow field, which is also strongly coupled to the volumetric heat generation. These considerations suggest that experimental and numerical modelling (including the option of coupled neutronics-thermal-hydraulics) would be needed to better understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built at BME NTI with the purpose of investigating the flow behavior inside the core region using particle image velocimetry. Not only the basic flow behavior inside the core region can be investigated but measurement data can also provide resource for the validation of computational fluid dynamics models, specific problems or phenomena (for example inlet geometry, optional internal structures, mixing) may be studied as well. Measurement results of steady state conditions will be presented with comparison of measurement data and results of numerical analyses. (author)

  5. 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

  6. Experimental determination of neutron capture cross sections of fast reactor structure materials integrated in intermediate energy spectra. Vol. 2: description of experimental structure

    International Nuclear Information System (INIS)

    A selection of technical documents is given concerning the experimental determination of the neutron capture cross-sections of fast reactor structural materials (Fe, Cr, Ni...) integrated over the intermediate energy spectra. The experimental structure project and modifications of the reactor RB2 for this experiment, together with criticality and safety calculations, are presented

  7. Three-dimensional simulation and experimental investigation of a novel biomass fast pyrolysis reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H.Y.; Shao, S.S.; Xiao, R.; Pan, Q.W.; Chen, R.; Zhang, J.B. [Southeast Univ., Nanjing (China). School of Energy and Environment

    2013-07-01

    A novel autothermal reactor, named internally interconnected fluidized beds (IIFB), was developed for biomass fast pyrolysis to produce liquid fuels and chemicals. The IIFB reactor includes a pyrolysis bed and a combustion bed to conduct biomass pyrolysis and char burning, respectively. In this study, numerical simulation and experimental studies on volume fraction of particles, solid circulation rate and pressure distribution of the IIFB are reported. The stable flow photographed from the simulations coincides with that in the experiments at the same operating conditions. At the same height, the velocity of gas is twice as larger as the velocity of solid, which is favorable for catalytic reactions. The particles move up unsteadily in the draft tube, and yet they fall down with an almost constant velocity 0.07 m/s in the dipleg. The pressure in the fluidization region is higher than that in the spouted region at H=10mm and it shows an opposite pressure distribution. It is also observed that the experimental value of pressure is in well agreement with that obtained from simulations on the bottom, and yet it shows very different characteristics on the two outlets. Simulation results show that solid circulation rate at different cross-sections converged to 110kg/h which is in well agreement with experimental data of 104.5kg/h.

  8. Development of Experimental System for Material Compatibility Test for Ultra-long Cycle Fast Reactor (UCFR)

    International Nuclear Information System (INIS)

    Sodium is a candidate for fast reactor coolants that has been believed to have favorable compatibility with structural materials. However, recent studies showed results which need for a more careful attention at this previous belief. For prolonging the service life time of cladding and structural materials in contact with liquid sodium, more detail analysis methods are needed to examine this material compatibility issue with sodium. As a candidate of liquid metals coolants of Ultra-long Cycle Fast Reactor (UCFR), the compatibility of sodium with cladding materials has to be investigated in detail with long term exposure time. It is known that sodium promotes corrosion in two ways. One is corrosion produced by dissolution of alloy elements into sodium and the other is corrosion produced through a chemical reaction with impurities in sodium (especially, dissolved oxygen). The use of the technique of impedance spectroscopy to measure the electrical impedance response of any oxide layers may be a good experimental tool to this monitoring system. The motivation of current study is to investigate the relationship between the electrochemical behaviors of oxide scales on martensitic and austenitic steels and their corrosion rates in liquid sodium

  9. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    International Nuclear Information System (INIS)

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U235 are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained

  10. Fast reactor database

    International Nuclear Information System (INIS)

    This publication contains detailed data on liquid metal cooled fast reactors (LMFRs), specifically plant parameters and design details. Each LMFR power plant is characterized by about 400 parameters, by design data and by relevant materials. The report provides general and detailed design characteristics including structural materials, data on experimental, demonstration, prototype and commercial size LMFRs. The focus is on practical issues that are useful to engineers, scientists, managers and university students and professors. The report includes updated information contained in IAEA previous publications on LMFR plant parameters: IWGRF/51 (1985) and IWGFR/80 (1991) and reflects experience gained from two consultants meetings held in Vienna (1993,1994). This compilation of data was produced by members of the IAEA International Working Group on Fast Reactors (IWGFR)

  11. Replacement of secondary heat transport system components in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    A recently completed major upgrade of the JOYO experimental sodium-cooled fast reactor, to the MK-III design, increased its irradiation capability. One major change was a 40% increase in thermal power to 140 MWt, which necessitated the replacement of the cooling system. Major challenges in the replacement of secondary components were control of impurity ingress and assurance of welding integrity. Damage to existing systems was avoided during replacement operations by taking measures to prevent ingress of air into the sodium systems. The long exposure of the used pipes made of ferritic low-alloy steel to hot sodium was a concern because previous research showed that this material changes its mechanical property in sodium hotter than 673 K. Used pipe, heat transfer tubes and welds were subjected to material tests. These tests did not show notable material problems. The replacement of components was completed without major troubles, demonstrating the effectiveness of the methods used. (authors)

  12. Utilization of the experimental reactor Osiris for the study and the development of fuels of the fast neutron reactor type

    International Nuclear Information System (INIS)

    Nuclear fuel tests for the fast neutron reactor type have been carried out at the Osiris reactor: thermal study of (U,Pu)O2 oxide by measurement with thermocouples in the core of the fuel pellet; study of the effects of power cycling on nuclear fuel; study of the mechanical interactions between oxide and cladding by measurement of the cladding deformation during irradiation

  13. Upgrade of Cooling System Heat Removal Capacity of the Experimental Fast Reactor JOYO

    International Nuclear Information System (INIS)

    The purpose of the MK-III program is to upgrade the irradiation capability of the liquid sodium-cooled experimental fast reactor JOYO. As a result, the neutron flux density of the core was increased, and the reactor thermal power was increased to 140 MW(thermal) from the originally designed 100 MW(thermal). To accommodate the increased thermal power, the flow rates of sodium coolant in the primary and secondary systems were increased by 20 and 10%, respectively. Also, all intermediate heat exchangers and dump heat exchangers were replaced with new ones. The replacement of these large sodium components was carried out over an [approximately]1-yr period with both fuel and molten sodium still in the reactor vessel (RV).Major challenges in the replacement were the control of impurity ingress to existing systems and protection from radiation exposure in the high-dose-rate regions inside the containment vessel. During the replacement, the seal bag method, impurity concentration monitoring of cover gas, and low-pressure control of cover gas were applied to prevent damage to existing components and systems, such as the RV, fuel subassemblies, sodium piping, and tanks. The measures taken to reduce the radiation exposure were a lowering of the surrounding dose rate through the use of temporary shielding, shortening of the operation time near the high-dose-rate area by first doing thorough training, and the employment of protection equipment to avoid contamination. The replacement of components was completed without major trouble, and methods applied for the replacement proved to be effective in the operation and maintenance of sodium-cooled reactors

  14. Student internship program using the experimental fast reactor Joyo and related facilities

    International Nuclear Information System (INIS)

    The student training courses using the experimental fast reactor Joyo of the Japan Atomic Energy Agency (JAEA) and related facilities have been initiated based on the JAEA's mission to contribute to the human resources development program of the Japanese Ministry of Education, Culture, Sports, Science and Technology (MEXT) and Ministry of Economy, Trade and Industry (METI). The development of the student training courses was also strongly supported by the faculty of nuclear engineering of domestic universities in two reasons: one is that the nuclear related curriculum has recently been reduced due to the trend of decreasing interest by the younger generation in nuclear research and industry, and the other reason is that the aging research reactors and nuclear facilities owned by the universities are very difficult to keep operating. Considering this situation, JAEA decided to cooperate with the universities in developing the student training course. The experimental fast reactor Joyo of JAEA is a sodium cooled fast reactor with plutonium- uranium mixed oxide (MOX) fuel, which has two primary sodium loops, two secondary loops, and an auxiliary system. An intermediate heat exchanger (IHX) separates radioactive sodium in the primary system from non-radioactive sodium in the secondary system. The secondary sodium loop transports the reactor heat from the IHX to the air-cooled dump heat exchanger (DHX). Joyo has a full-scope type core and plant simulator, which duplicated all the main control panels located in the Joyo central control room. The simulator enables to offer a real time simulation of the plant behaviors under normal and abnormal conditions by applying the plant dynamic analysis code Minir-N2 and the same interlock system as the Joyo reactor system. The sodium analysis facility is located apart from Joyo complex to primarily conduct impurity measurement of Joyo cooling system. These data were measured by chemical analysis, gas chromatography, beta

  15. Major accident analyses for experimental zero-power fast reactor assemblies

    International Nuclear Information System (INIS)

    A study has been made of the possibility, mechanism, and consequence of melt-down and other major nuclear accidents for a ZPR-III type experimental zero-power fast reactor of the two-half type. This study has been supplemented by an evaluation of the importance of the Doppler effect for a wide range of nuclear reactor assemblies for such a reactor. A melt-down event is highly improbable because of the restricted sequence of events which must be postulated. A discussion of the mechanism of the collapse is followed by the results of coupled neutronics-hydrodynamic s calculations for two zero-power assemblies. A 1200-l core has been examined because it represents a relatively large reactor of common core composition. A smaller core with a high-void fraction has been examined as a potentially more dangerous system. Very different time-wise behaviour has been found for the two systems. For sharp accidents in zero-power assemblies, the U235-atoms, separated as plates of enriched uranium, will heat very rapidly while the remainder of the core remains essentially cold, so that a gas of U235-vapour will provide the disassembly pressure. The adaption of the neutronics-hydrodynamic s code AX-I to the use of a Van der Waals gas is described. Another important change in the equation of state used in the code is to employ a Mie-Griineisen type equation derivable from solid state theory. This change provides a more satisfactory way to evaluate the pressure term for cores of variable composition. Because the highly enriched U235 plates of a zero-power assembly will heat much more rapidly than the depleted uranium plates, the possibility of a net positive Doppler effect is much larger for an experimental assembly than for the equivalent power breeder reactor. This hazard has been examined for a range of possible assemblies. These calculations indicate that the Doppler coefficient for a zero-power assembly does not become important as a hazard until one approaches systems with the

  16. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  17. 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)

  18. Measurement and evaluation of Corrosion Products deposition distribution in the Experimental Fast Reactor JOYO

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Takafumi; Sumino, Kozo [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Masui, Tomohiko; Saikawa, Takuya

    1997-12-01

    The Corrosion Product (CP) is the major radiation source in the primary cooling system of an LMFBR plant. It is important to characterize and predict the CP behavior to reduce the personnel exposure dose due to CP deposition. The CP measurement was carried out in the Experimental Fast Reactor JOYO during the 11th annual inspection period when the accumulated reactor thermal power reached about 143 GWd. The CP deposition density was measured using a pure germanium detector. The plastic scintillation fiber (PSF) was applied for the gamma-ray dose rate distribution measurement and compared with the thermoluminescence dosimeter (TLD). The major results obtained by the CP measurements in JOYO are the follows: (1) The major CP nuclides deposited in the primary cooling system are {sup 54}Mn and {sup 60}Co. {sup 54}Mn is the dominant isotope and it tends to deposit in the cold leg region. On the other hand, {sup 60}Co deposits mainly in the hot leg region. The deposition density of {sup 54}Mn is about seven times as much as that of {sup 60}Co in the cold leg region and twice in the hot leg region. (2) The deposition densities of {sup 54}Mn and {sup 60}Co, and the gamma-dose rate were decreased from the last data in the previous annual inspection period mainly due to the short operation time and the longer cooling time. (3) The continuous gamma-ray dose rate distribution up to 10m can be measured by using the PSF in a few minutes. The PSF is suitable to measure the gamma-ray dose rate distribution in the maintenance work area where it is narrow and the mixture of gamma-ray sources from primary pipings and components. The data base of detailed gamma-ray dose rate distribution was greatly extended by the PSF. (author)

  19. 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

  20. 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.

  1. Commercialization of fast reactors

    International Nuclear Information System (INIS)

    Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)

  2. Joyo experimental reactor tour

    International Nuclear Information System (INIS)

    JAEA cooperation in remote monitoring focuses on the Joyo Experimental Reactor at the O'arai Research and Development Center. Joyo performs irradiation of test fuels to support development of the fast reactor cycle in Japan, both in international cooperation and in support of the Monju fast reactor, which is now undergoing reconstruction. The tour included an introduction at the model, a visit to the control room, entry into the containment vessel, and viewing of remote monitoring equipment in the Fresh Fuel Storage and at one of the Spent Fuel Ponds. (author)

  3. Experimental research of local hydrodynamic characteristics of fast reactor fuel assembly peripheral zone. 4

    International Nuclear Information System (INIS)

    Measurements were made of shear stress distribution and the velocity field of an aerodynamic model of the fast breeder reactor fuel assembly periphery. The effect was studied of a 50% disturbance of the geometry of a corner rod in a fuel assembly as against normal geometry. The coefficient of friction in the channel was assessed in dependence on the Reynolds number. The distribution of shear stresses in the walls of the fuel assembly and on spacers is graphically represented. (M.D.)

  4. Experimental research of local hydrodynamic characteristics of fast reactor fuel assembly peripheral zone. 5

    International Nuclear Information System (INIS)

    The results are presented of measurements of shear stress distribution and velocity fields on an aerodynamical model of a fast breeder reactor fuel subassembly periphery. The peripheral configuration is disturbed by two side rods displaced towards one rod of the second row. These three rods form a symmetrical cell with a relative pitch of 1.0325. A comparison is made of the results and results with nominal configuration. (author). 75 figs., 10 tabs., 13 refs

  5. Fast reactor operating experience

    International Nuclear Information System (INIS)

    At the beginning of electricity generation from nuclear power there was the breeder, which fulfilled its duty in a number of smaller test and experimental reactors within national programs. Over the years, some of those reactors have attained impressive availabilities, while others have helped to improve our knowledge by the negative results they contributed. Worldwide a decisive step was taken by the mid- to late sixties in the planning and construction of medium sized demonstration fast breeder power plants (250 to 350 MW). In the Federal Republik of Germany, this step is taken belatedly in building the SNR-300. BN-350 in the USSR, Phenix in France, and PFR in the United Kingdom have now been in operation for some ten years. Over that period, valuable experience has been accumulated in sodium technology. The operating behavior of all components and systems working in sodium is called excellent; the hazards associated with sodium, the fire hazard in particular, thus often seem to be greatly overrated. Leakages have been brought under control. It has always been possible so far to trace them back to systemic faults produced in the welding process. The ability of fast sodium cooled reactors to produce more nuclear fuel than they consume has been demonstrated in Phenix, whose breeding ration has been measured to be 1.16. The first true large breeder, Super Phenix in France, is to be commissioned already in 1985. In building another three breeder power plants the European partners in an association hope to achieve the commercial breakthrough of the breeder line. (orig.)

  6. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  7. Experimental fast reactor Joyo emergency operation act in Northeastern Japan Earthquake 3.11

    International Nuclear Information System (INIS)

    Experimental fast reactor 'Joyo' under facility's periodic inspection received damage at its power incoming unit due to the 2011 off the Pacific Coast of Tohoku Earthquake, which incurred loss-of-offsite-power. Immediately, two emergency diesel generators (D/G) automatically started, which supplied emergency system power. During eight days before the power incoming unit got provisional restoration, power supply through D/G continued. During this period, emergency measures for fuel and cooling water securement for D/G were taken, which forced unexperienced long-term load operation for D/G. This paper reports the change of plant conditions of Joyo, as well as the measures taken for keeping fuel and cooling water for continuing the operation of D/G. Since the main machines of the primary system and secondary system were functioning normally after the earthquake, the plant could be maintained in a stable state. As for the fuel for D/G, reduction in fuel consumption based on load suppression was implemented, and the cooperation of trading firms, related facilities, and local suppliers secured the fuel that supported the long-term operation of D/G. As for the cooling water, the cooperation of a self-defense fire brigade allowed to secure the required amount by utilizing the reserved water in the water tank for fire prevention. From these series of experiences of handling the plant, it was possible to extract the future challenges against the time when a massive earthquake or prolonged power failure has occurred. (A.O.)

  8. Experimental studies on natural circulation decay heat removal in Japan Sodium Cooled Fast Reactor (JSFR)

    International Nuclear Information System (INIS)

    Fully natural circulation system is adopted in a decay heat removal system (DHRS) of the designs of Japan Sodium Cooled Fast Reactor (JSFR). Several investigations of experiments and simulation methods on this DHRS were performed. Water experiments were carried out for the primary heat transportation system including a reactor vessel and heat exchangers of DHRS using a 1/10 model. As for the DHRS loop, sodium experiments were carried out, especially for a heat exchanger installed in an Intermediate Heat Exchanger (IHX). Here, several results of the sodium experiments were described. Transient characteristics during the start up in the air system of the air cooler, secondary loop of DHRS, and the primary loop were examined by the sodium experiments. Smooth increases of natural circulation flow rates in all systems of air and sodium were confirmed. Verifications of numerical simulation methods are planned based on the water and sodium experiments in this investigation plan. (author)

  9. 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

  10. Overview on New Research Reactors in China

    International Nuclear Information System (INIS)

    In China, 2 research reactors are now under construction. Correspondingly, this paper consists of 2 parts. Part 1 will focus on China Advanced Research Reactor (CARR), the reactor characteristics, utilization, safety related systems and other main systems will be described in this part. Part 2 will focus on China Experiment Fast Reactor(CEFR), the general design and the safety features in particular will be illustrated in this part. (author)

  11. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. 2. Replacement of upper core structure

    International Nuclear Information System (INIS)

    In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of MARICO-2 (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS) in 2007. As a part of the restoration work, UCS replacement was begun at March 24, 2014 and was completed at December 17. In-vessel repair (including observation) for sodium-cooled fast reactors (SFRs) is distinct from that for light water reactors and necessitates independent development. Application of developed in-vessel repair techniques to operation and maintenance of SFRs enhanced their safety and integrity. There is little UCS replacement experience in the world and this experience and insights, which were accumulated in the replacement work of in-vessel large structure (UCS) used for more than 30 years, are expected to improve the in-vessel repair techniques in SFRs. (author)

  12. Demonstration test of the holding stability of the self actuated shutdown system in the experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large scale fast breeder reactor (FBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the stability of SASS needs to be confirmed under the actual reactor-operational environment with high temperature, high neutron flux, and flowing sodium in order to ensure the high plant availability factor. For this purpose, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor JOYO MK-III. As a result of this study, the rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation. (author)

  13. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

  14. Preliminary Conceptual Design for a Multipurpose Experimental Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    A conceptual nuclear design was done for a 100MWe multi-purpose sodium cooled fast reactor core for both research and experiment as a pre-feasibility study. Challenging goals for a small sized core were set up. Cycle length should be longer than 4 month with low enriched uranium-only fuels in a U-Zr metal alloy form. Overall design features are chosen from previous design concept - advanced burner test reactor (ABTR). Number of fuel assemblies and reflector assemblies were increased from 54 to 66 and from 78 to 207 respectively, resulting in increase of core radius to 180cm. In the reflector zone, there was enough space to install three independent fuel test loops in which different type of coolants and fuels might be tested. Fuel pin size was selected to be 0.737cm through a parametric study with database for U-Zr metal fuels tested in EBR-II. Core height was optimized to be 87cm after a parametric study on core size. Target flux levels at both fuel test assemblies and material test assemblies were 3.0×1015n/cm2-sec. Most of performance parameters as well as safety parameters were satisfied with goals except a flux level at the MTA. Overall safety aspects of the reactor is better than any other previous SFR design concepts

  15. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

    International Nuclear Information System (INIS)

    In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-α techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence

  16. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    International Nuclear Information System (INIS)

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

  17. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  18. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarises the fast reactor research carried out at the Netherlands Energy Research Centre during the year 1981. The neutron and fission product cross sections of various isotopes have been evaluated. In the fuel performance programme, some preliminary results are given and irradiation facilities described. Creep experiments on various stainless steel components are reported

  19. Modelling of Multi-Physics Phenomena in Fast Reactor Design: Safety and Experimental Validation

    International Nuclear Information System (INIS)

    The paper provides a cursory look at current approaches in numerical modelling and simulation of typical multi-physics phenomena of concern relevant to the sodium cooled fast reactor design and safety. Emphasis is placed on the methods that are in practice and their verification and validation programmes, including for those of fluid–structure thermal interactions due to thermal striping, thermodynamics of sodium–water chemical reactions, multi-component and multi-phase flows in the fuel degradation and core meltdown phases. Several of the numerical simulations of these phenomena are shown with verification and validation programs that employ not only separate effect small scale experiments of clean geometry but also for large scale integral tests or mock-up experiments. The last part of this paper will be spent on discussions on a more quantitative validation basis with identification of errors and/or uncertainties based on the Bayesian rule. (author)

  20. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  1. Emergency procedure control for station blackout in the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    After the 2011 off the Pacific Coast of Tohoku Earthquake and the Fukushima Daiichi Nuclear Power Station accident on March 11, 2011, the safety ensuring of reactor facilities at the time of power source function loss has been required more than ever. The Nuclear Regulatory Commission was established in September 2013, and the study on safety standards has been promoted in the framework of new regulations. In Joyo, under these circumstances, voluntary review and countermeasures are advancing in order to improve corresponding measures against power source function loss. As one of such efforts, guidelines in case of station blackout were prepared. In the sodium-cooled fast reactor 'Joyo', it is possible as plant characteristics that the core decay heat can be removed by natural circulation cooling even when the forced cooling function has been lost. Moreover, since power supply from an uninterruptible power supply system is possible even after station blackout, necessary monitoring functions can be maintained, However, it is necessary to consider the response measures for the discharge of storage battery. This paper reports the contents of study on corresponding measures for the incident from the start to end, as well as the contents of the preparation for the equipment for temporary power supply etc., and the training that was carried out on the assumption of the same incident. (A.O.)

  2. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  3. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  4. Gas-cooled fast reactors. Motivation and presentation of the ENIGMA program in the MASURCA experimental critical facility

    International Nuclear Information System (INIS)

    This paper describes a new experimental physics program in support of gas cooled fast reactor (GCFR) design studies, called ENIGMA, to be performed in the MASURCA critical facility at CEA-Cadarache, France. The prospective GCFR design studies at CEA are presented, as well as the specific neutronics features needing an extension of the validation of calculation tools and nuclear data. The relevant existing experiments are briefly reviewed and the need for new experimental data is pointed out. The first phase of the proposed new experiments includes a reference core with a representative spectrum, and a series of central core substitutions involving spectrum shifts, streaming studies, low-grade Pu substitutions, innovative material (Si, Zr) substitutions. Reflector substitution zones will include elements foreseen for the reflectors (Si, Zr, C). Subsequent phases will involve larger amounts of low-grade Pu or innovative materials, and configurations representative of experimental and demonstration GCFRs. (author)

  5. Experience of primary cooling system modification to increase heat removal capability in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    The purpose of the MK-III program is to upgrade the irradiation capability of the JOYO experimental sodium-cooled fast reactor. As a result, the neutron flux density of the core was increased and the reactor thermal power was increased to 140 MWt from the originally designed 100 MWt. To accommodate the increased thermal power, the flow rates of sodium coolant in the primary and secondary systems were increased by 20 % and 10 %, respectively. Also, all intermediate heat exchangers and dump heat exchangers were replaced with new ones. During this replacement, molten sodium and fuel were retained in the reactor vessel. Consequently, the primary cooling system and cover gas boundaries had to be maintained to prevent impurity ingress to the sodium system. During the replacement, the seal bag method, impurity concentration monitoring of cover gas, and low-pressure control of cover gas were applied to prevent damage to existing components and systems. A roller cutter was used for cutting large diameter pipes to prevent ingress of cuttings. The measures taken to reduce the radiation exposure were a lowering of the surrounding dose rate through the use of temporary shielding, shortening of the operation time near the high dose rate area by first doing thorough training, and the employment of protection equipment to avoid contamination. The replacement of components was completed without major trouble, and methods applied for the replacement proved to be effective in the operation and maintenance of sodium cooled reactors. (author)

  6. Study on improvement of core management and irradiations field characterization methods of the experimental fast reactor Joyo (Thesis)

    International Nuclear Information System (INIS)

    This thesis describes the research study to develop the core management method and irradiation field characterization method of the experimental fast reactor Joyo. Improvements of the methods through comparison with measured data from the reactor core physics performance tests of Joyo and post irradiation examination (PIE) of tests conducted in the Joyo irradiation test facility complex are also described. There are eight chapters. Chapter 1 describes the objectives of this study, along with a brief history of the Joyo test reactor and an explanation of the role and importance of developing the sodium cooled fast breeder reactor (FBR) in Japan from the view point of providing the future energy source. Chapter 2 explains the core management method of the Joyo Mark-II irradiation core, which had been modified from the first Mark-I breeder core. The core management method modifications of Joyo included changing the refueling scheme by employing an in-out fuel shuffling method and re-examination of the thermal design margin of the driver fuel by reducing the hot spot factor based on the evaluation of the Joyo Mark-II core and plant performance tests. Chapter 3 describes the development of improved methods for evaluating the neutron and gamma flux distributions by including energy spectrum information in order to meet the requirements for their accuracy. These developments included modifying the analytical method and developing the new neutron dosimetry method of helium accumulation fluence monitor (HAFM). These improvements were validated by comparison with the measured reaction rates obtained by the conventional multiple foil activation method. Chapters 4 and 5 describe the design of the upgrade of the Joyo core and cooling system, called the Mark-III project, in order to increase the neutron flux 1.3 times higher than the original design maximum of the Joyo Mark-II core. The modified Mark-III core and plant performance test evaluations that were used to validate the

  7. 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)

  8. 1995 benchmark data based on experimental results from the prototype fast reactor at Dounreay

    International Nuclear Information System (INIS)

    During the final year of operation of the Prototype Fast Reactor (PFR) at Dounreay and during the first 3 months following its closure on March 31 1994, a series of experiments was conducted to obtain data on the performance of leak detection systems. In June 1994, a series of injections of argon, hydrogen and steam was performed in Evaporator 3. Two injection locations were studied: one within the tube bundle region and one in the flowing sodium interspace between the tube bundle wrapper and the steam generator shell. During 113 injections, acoustic noise measurements were made at 7 transducer locations, spatially distributed on the shell of the Evaporator. These transducers were mounted on waveguides, welded to the shell. Data from two argon injections, two steam injections and one hydrogen injection in the tube bundles and from 1 argon, 1 steam and 1 hydrogen injection in the sodium interspace, recorded at 4 transducer locations, were selected for the IAEA 1995 Benchmark exercise. The plant state during these injections was such that the acoustic background noise was lower than at full power operating conditions. It was agreed at the meeting to discuss the 1994 Benchmark results and to agree the 1995 Benchmark data, that it would be preferable not to mix these injection signals with full power background noise, but to include some separate full power background noise data for the Evaporator and for a Superheater. Accordingly, full power data recorded at two transducer locations in each unit have been included with the injection data. 11 figs, 10 tabs

  9. Monte Carlo simulation analysis of integral data measured in the SCK-CEN/ENEA experimental campaign on the TAPIRO fast reactor. Experimental and calculated data comparison

    Energy Technology Data Exchange (ETDEWEB)

    Burgio, N., E-mail: nunzio.burgio@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), C.R. Casaccia Via Anguillarese 301, 00123 Rome (Italy); Cretara, L., E-mail: luca.cretara@uniroma1.it [DIAEE – Sapienza University of Rome, Corso Vittorio Emanuele II 244, 00186 Rome (Italy); Frullini, M., E-mail: massimo.frullini@uniroma1.it [DIAEE – Sapienza University of Rome, Corso Vittorio Emanuele II 244, 00186 Rome (Italy); Gandini, A., E-mail: augusto.gandini@uniroma1.it [DIAEE – Sapienza University of Rome, Corso Vittorio Emanuele II 244, 00186 Rome (Italy); Peluso, V., E-mail: vincenzogiuseppe.peluso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), Via Martiri di monte Sole 4, 40129 Bologna (Italy); Santagata, A., E-mail: alfonso.santagata@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), C.R. Casaccia, Via Anguillarese 301, 00123 Rome (Italy)

    2014-07-01

    Highlights: • We develop a MCNPX model of the TAPIRO fast research reactor. • The model has been tested against the result of a late experimental champaign finding on overall agreement. • The source of uncertainties in the nuclear data and in the model assumptions has been discussed. • The model is sufficiently accurate to design irradiation experiment in support to R and D activities on LFR and ADS systems. - Abstract: After Fukushima events, the Italian nuclear program has been redefined leaving space only to activities related to Generation IV nuclear systems. Accordingly with this renewed national scenario, TAPIRO fast reactor facility is gaining a relatively major strategic role. A program is in fact being proposed to host in TAPIRO benchmark experimental activities related to the development of Lead fast reactor and Accelerator Driven Systems. A first step of this program would consist on the validation of neutronic codes, cross section data and reactor models to be adopted for its analysis. Along this line in this work the results of a simulation study has been made relevant to the measurements performed in the SCK-CEN/ENEA experimental campaign carried out in the 1980–1986 period. The calculations have been made using the Monte Carlo MCNPX 2.7.0 Code. In this article the main results are presented and discussed, with particular emphasis on the uncertainties, relevant both to nuclear data and the model layout. The results of this simulation study indicate in particular that TAPIRO's MCNPX model is adequate for the optimization of set-ups of perspective neutron irradiation experiments, this allowing cuts in costs and development time.

  10. Monte Carlo simulation analysis of integral data measured in the SCK-CEN/ENEA experimental campaign on the TAPIRO fast reactor. Experimental and calculated data comparison

    International Nuclear Information System (INIS)

    Highlights: • We develop a MCNPX model of the TAPIRO fast research reactor. • The model has been tested against the result of a late experimental champaign finding on overall agreement. • The source of uncertainties in the nuclear data and in the model assumptions has been discussed. • The model is sufficiently accurate to design irradiation experiment in support to R and D activities on LFR and ADS systems. - Abstract: After Fukushima events, the Italian nuclear program has been redefined leaving space only to activities related to Generation IV nuclear systems. Accordingly with this renewed national scenario, TAPIRO fast reactor facility is gaining a relatively major strategic role. A program is in fact being proposed to host in TAPIRO benchmark experimental activities related to the development of Lead fast reactor and Accelerator Driven Systems. A first step of this program would consist on the validation of neutronic codes, cross section data and reactor models to be adopted for its analysis. Along this line in this work the results of a simulation study has been made relevant to the measurements performed in the SCK-CEN/ENEA experimental campaign carried out in the 1980–1986 period. The calculations have been made using the Monte Carlo MCNPX 2.7.0 Code. In this article the main results are presented and discussed, with particular emphasis on the uncertainties, relevant both to nuclear data and the model layout. The results of this simulation study indicate in particular that TAPIRO's MCNPX model is adequate for the optimization of set-ups of perspective neutron irradiation experiments, this allowing cuts in costs and development time

  11. Experimental and numerical investigations on roof slab of a pool type sodium cooled fast reactor based on model studies

    International Nuclear Information System (INIS)

    Highlights: • The structural integrity of the roof slab of 500 MWe sodium cooled fast reactor is predicted under static loading conditions. • A scaling down approach was adopted to reduce the cost of experimentation. • Experiments on 1/12th scaled down model of the roof slab made of Perspex material are carried out and test data is compared with the results of numerical simulation. • The investigations carried out on the model demonstrate the robustness of the analysis as well as raise the confidence on the structural integrity of the roof slab. - Abstract: The objective of the work is to predict the structural integrity of the roof slab of 500 MWe sodium cooled fast reactor (SFR) under static loading conditions. The roof slab is an annular box type structure consisting of top and bottom plates with connecting stiffeners and has been designed to support various components such as control plug, pumps and heat exchangers entering into the main vessel of the reactor. The net static load acting on it is about 3800 t under normal condition and about 5000 t under design basis earthquake. Experiments on 1/12th scaled down model of the roof slab are carried out and test data is compared with the results of numerical simulation by finite element analysis. The numerical as well as experimental investigations carried out on the model demonstrate the robustness of assumptions made for the analysis carried out towards respecting design code limits as well as raise the confidence on the structural integrity of the roof slab

  12. A review of theoretical and experimental studies underlying the thermal-hydraulic design of fast reactor fuel elements

    International Nuclear Information System (INIS)

    The economic performance of fast reactors is closely linked to the achievable burn-up of heavy atoms, that is to the endurance life of the fuel pins. The safety case must also be concerned with the integrity of the cladding, since this is the primary containment envelope for fission products. It is thus important to ensure that cladding temperatures during reactor operation are limited to levels which incur no serious impairment of mechanical properties. The function of thermal-hydraulic analysis is to provide fuel element designers with the means of achieving this objective. This paper reviews the theoretical approaches which have been developed and applied in the UK in the design of LMFBR fuel and breeder sub-assemblies, control rods and experimental clusters. It also presents results of experimental studies undertaken to develop a better understanding of coolant flow distribution and mixing problems in these components, and to provide essential data for computer codes. Problem areas in this field are highlighted, particularly the difficulties arising due to irradiation induced distortions. Reference is made to the experimental and theoretical developments which are in progress, or may be required, to provide adequate predictions of fuel pin temperatures at high burn-up. (author)

  13. On fast reactor kinetics studies

    Energy Technology Data Exchange (ETDEWEB)

    Seleznev, E. F.; Belov, A. A. [Nuclear Safety Inst. of the Russian Academy of Sciences IBRAE (Russian Federation); Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F. [Inst. for Physics and Power Engineering IPPE (Russian Federation)

    2012-07-01

    The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

  14. 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.

  15. Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997

  16. Fast reactors and nonproliferation

    International Nuclear Information System (INIS)

    1.Three aspects of nonproliferation relevant to nuclear power are: Pu buildup in NPP spent fuel cooling ponds (∼ 104 t in case of consumption of ∼ 107 t cheap uranium). Danger of illegal radiochemical extraction of Pu for weapons production; Pu extraction from NPP fuel at the plants available in nuclear countries, its burning along with weapon-grade Pu in NPP reactors or in special-purpose burners; increased hazard of nuclear weapons sprawl with breeders and closed fuel cycle technology spreading all over the world. 2.The latter is one of major obstacles to creation of large-scale nuclear power. 3.Nuclear power of the first stage using 235 U will be able to meet the demands of certain fuel-deficient countries and regions, replacing ∼ 5-10% of conventional fuels in the global consumption for a number of decades. 4.Fast reactors of the first generation and the currently employed fuel technology are far from exhausting their potential for solving economic problems and meeting the challenges of safety, radioactive waste and nonproliferation. Development of large-scale nuclear power will become an option accepted by society for solving energy problems in the following century, provided a breeder technology is elaborated and demonstrated in the next 15-20 years, which would comply with the totality of the following requirement: full internal Pu breeding deterministic elimination of severe accidents involving fuel damage and high radioactivity releases: fast runaway, loss of coolant, fires, steam and hydrogen explosions, etc.; reaching a balance between radioactive wastes disposed of and uranium mined in terms of radiation hazard; technology of closed fuel cycle preventing its use for Pu extraction and permitting physical protection from fuel thefts;economic competitiveness of nuclear power for most of countries and regions, i.e. primarily the cost of NPPs with fat reactors is to be below the cost of modern LWR plants, etc

  17. Experimental studies of similarity criteria for gas entrainment phenomena using water models of a fast reactor hot plenum

    International Nuclear Information System (INIS)

    Entrainment of the argon cover gas blanket into the bulk sodium flow of a fast reactor hot plenum is undesirable for a number of reasons, particularly if such gas is transported around the primary circuit. Possible entrainment mechanisms which can occur at the hot plenum free surface include waves, downward flows and more significantly, vortices produced by eddy shedding off structures. Experimental studies of such phenomena using sodium facilities are prohibitive both in terms of cost and technical difficulty. Unfortunately modelling laws for these entrainment mechanisms are not well defined and it has been difficult to define or quantify criteria by which the acceptability of the free surface of a reduced scale model, operating with simulant fluid, could be judged. This paper describes how resistive wave gauges may be used in conjunction with suitably designed data analysis techniques to characterise free surface fluctuations in terms of quantifiable parameters. (orig.)

  18. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling

  19. 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

  20. Scientific and technical conference Thermophysical experimental and calculating and theoretical studies to justify characteristics and safety of fast reactors. Thermophysics-2012. Book of abstracts

    International Nuclear Information System (INIS)

    The collection includes abstracts of reports of scientific and technical conference Thermophysics-2012 which has taken place on October 24-26, 2012 in Obninsk. In abstracts the following questions are considered: experimental and calculating and theoretical studies of thermal hydraulics of liquid-metal cooled fast reactors to justify their characteristics and safety; physico-chemical processes in the systems with liquid-metal coolants (LMC); physico-chemical characteristics and thermophysical properties of LMC; development of models, computational methods and calculational codes for simulating processes of of hydrodynamics, heat and mass transfer, including impurities mass transfer in the systems with LMC; methods and means for control of composition and condition of LMC in fast reactor circuits on impurities and purification from them; apparatuses, equipment and technological processes at the work with LMC taking into account the ecology, including fast reactors decommissioning; measuring techniques, sensors and devices for experimental studies of heat and mass transfer in the systems with LMC

  1. Design and Experimental Study for Development of Pb-Bi Cooled Direct Contact Boiling Water Small Fast Reactor (PBWFR)

    International Nuclear Information System (INIS)

    A design concept of Pb-Bi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. In the PBWFR, water is injected into hot Pb-Bi above the core, and direct contact boiling takes place in the chimney. The boiling two-phase flow in the chimney serves as a steam lift pump and a steam generator. A two-region core is designed. A decrease in reactivity was estimated to be 1.5 % dk/kk' for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The chimney, cyclone separators and chevron dryers, direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed. For the technical development of the PBWFR, experimental and analytical studies are performed for Pb-Bi direct contact boiling two-phase flow, steel corrosion in Pb-Bi flow, oxygen control and oxygen sensor, and removal of polonium contamination. (authors)

  2. United States fast reactor programme

    International Nuclear Information System (INIS)

    The fast reactor programs in USA deal with: EBR-II termination program; fast flux test facility; nuclear energy research initiative; accelerator transmutation of waste; and other non-DOE funded activities. These are concerned with preliminary concept development activities of a secure transportable autonomous reactor -liquid metal cooled

  3. Periodic safety review of the experimental fast reactor JOYO. Review of the activity for safety

    International Nuclear Information System (INIS)

    Periodic safety review (Review of the activity for safety) which consisted of 'Comprehensive evaluation of operation experience' and Incorporation of the latest technical knowledge' was carried out up to January 2005. 1. Comprehensive evaluation of operation experience. It was confirmed that the effectual activities for safety through the operation of JOYO were carried out in terms of (1) Operation management, (2) Maintenance management, (3) Fuel management, (4) Radiation management, (5) Radioactive waste management, (6) Emergency planning and (7) Feedback of incidents and failures. 2. Reflection of the latest technical knowledge. It was confirmed that the latest technical knowledge including regulation and guide line established by Nuclear Safety Commission of Japan until March 31st. 2003 were properly reflected in impressing the safety of the reactor. As a result, it was evaluated that the activity for safety was carried out effectually, and no additional measure was identified continual safe operation of the reactor. (author)

  4. Summary of power ascension test of experimental fast reactor 'JOYO' MK-I

    International Nuclear Information System (INIS)

    On April 24th, 1977, the initial criticality of JOYO was achieved and on July 5th, 1978, the reactor output reached rated power of 50 MW for the first time. The 75MW power ascension test was started in July, 1979, followed by two cycles of rated power operations, and the 100 hour nominal power continuous operation was completed in February, 1980. Through the tests for the core, plant it self, radiation shield and plant monitoring, the results proved satisfactory operation characteristics at 75MW. This report presents the summary of all the results obtained in the Test of MK-I core. (author)

  5. Simulation on primary coolant system of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    In this paper the thermal-hydraulic characteristics of the primary loop of China Experimental Fast Reactor (CEFR) are calculated and analyzed. A one-dimension, single-phase flow model is used to establish the system control equations. The single channel model is adopted in the reactor core, and a dynamic model of intermediate heat exchanger is built. At the same time, the property of sodium and flow and heat transfer correlations or models of sodium are collected and compiled. The discussion of the sensitivity of different flow and heat transfer correlations is given. The validation of the code developed in this paper shows that the code can be adopted to do some typical transient and accident analysis. The model and code presented in this paper can be used not only in the safety analysis of pool-type sodium cooled fast reactor, but also in the development of CEFR simulation platform. (author)

  6. Calculational methods, codes and results of calculational and experimental investigations of control rod worth in power fast reactors

    International Nuclear Information System (INIS)

    The paper aims to present the main physical principles for selection of design characteristics of the fast reactor control rods (CR) system. The brief analysis of problems of CR physical calculations is given. Four components are described for the correction to the control rod worth calculated by the routine method based on the few - group three - dimensional diffusion code (TRIGEX) in hexagonal geometry. Principle considerations are given for the choice of the original task discretization methods implemented in this code to minimize the total error. Brief information is given about methods and codes used for the evaluation of error components of control rod worths calculated in a standard way. The results of experimental and calculational investigations of control rod physical characteristics are presented. These results were obtained at BFS critical assemblies simulating LMFBR cores. The investigations have been carried out for different types of core configurations. The experimental and calculated values are given on the distortion of power distribution due to the control rod insertion in the core. (author). 51 refs, 9 figs, 5 tabs

  7. Experimental and calculational study of temperature distributions in deformed model fuel assemblies of fast reactors

    International Nuclear Information System (INIS)

    Experimental and calculational data tastify to absence of temperature nonuniformity stabilization in fuel assembly peripheral area. The effect of fuel lattice deformation on the fuel assembly temperature field at shroud crushing in the core centre is demonstrated. 17 refs.; 21 figs

  8. Recycle strategies for fast reactors and related fuel cycle technologies

    International Nuclear Information System (INIS)

    . Although the U.S. returned to R and Ds on recycling in the 2000s, it has made a major shift from fast reactor cycle related and specific R and Ds to rather a long-term, science-based research program in accordance with the establishment of a new administration in January, 2009. Russia, China and India have promoted R and Ds independently depending on their own national conditions and nuclear energy policy aimed at completion of the closed cycle using fast reactors. In China and India, using MOX fuel at the moment, a shift to metal fuels is expected as countermeasures for the growth of energy demand and supply in the future. Further, Korea addresses R and Ds on metal-fueled fast reactors and pyro-processing systems, which have a relatively high proliferation resistance. Introduction of fast reactor cycle systems will be carried out independently by each country following the national conditions and nuclear energy policy. It should be then considered important to have globally common consensus relating to safety philosophy, concepts of proliferation resistance, TRU burnup and recycling and so on. 3. Multinational cooperation on future nuclear systems Twelve countries and one international organization have joined GIF (Generation IV International Forum) aiming to share R and Ds and tasks on Generation IV reactor systems (six reactor systems) in which five are fast reactors, and they have been promoting cooperative development and information exchanges in the framework of GIF. A common ground of understanding GIF has been nurtured in terms of its sustainability (efficient use of resources, minimization and management of nuclear wastes), economic efficiency, safety/reliability and proliferation resistance, and its technology goals have also been clearly defined. INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) intends to help to ensure that nuclear energy is available in the 21st century in a sustainable manner, led by IAEA and engaging twenty nine

  9. Experimental study of temperature field at fast reactor subassembly exit under drastic changes of coolant temperature

    International Nuclear Information System (INIS)

    Failure conditions due to dangerous increasing in power or flow rate drop are the most hazardous in terms of the rise of thermal stresses. Initial rise in temperature may run to 100 C and more. Sodium temperature at the subassembly inlet is varied according to definite time constant which is equal to fuel pin time constant (about 2 sec), that is below the time constant for massive part of subassembly head (4-10 sec). Thus, variations in sodium temperature are, for subassembly head, almost momentary and bring about maximal thermal stresses. Experiments on transient temperature behavior in subassembly head under thermal impact conditions have been performed on the model. Magnitude of temperature has been measured in two cross sections by chromel-alumel thermocouples bond in the middle of the wall, at its outer surface and in the coolant flow for distance of 3 mm from the wall. To measure temperature difference between middle of the wall and its surface fast differential thermocouples chromel-sodium-potassium have been used

  10. Fast reactor research activities in Brazil

    International Nuclear Information System (INIS)

    Fast reactor activities in Brazil have the objective of establishing a consistent knowledge basis which can serve as a support for a future transitions to the activities more directly related to design, construction and operation of an experimental fast reactor, although its materialization is still far from being decided. Due to the present economic difficulties and uncertainties, the program is modest and all efforts have been directed towards its consolidation, based on the understanding that this class of reactors will play an important role in the future and Brazil needs to be minimally prepared. The text describes the present status of those activities, emphasizing the main progress made in 1996. (author)

  11. Fast reactors and nuclear nonproliferation

    International Nuclear Information System (INIS)

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (author)

  12. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  13. Interfacial effects in fast reactors

    International Nuclear Information System (INIS)

    The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed to measure U-238 capture rates near th blanket--reflector interface in the MIT Blanket Test Facility. Prior MIT experiments on a thorium--uranium interface in a blanket assembly were also reanalyzed. Extremely localized fertile capture rate increases of on the order of 50% were measured immediately at the interfaces relative to extrapolation of asymptotic interior traverses, and relative to state-of-the-art (LIB-IV, SPHINX, ANISN/2DB) calculations which employ infinite-medium self-shielding throughout a given zone. A method was developed to compute a spatially varying background scattering cross section per absorber nucleus which takes into account both homogeneous and heterogeneous effects on the interface flux transient

  14. Probabilistic safety assessment method for sodium fire of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    The sodium fire is a typical and distinctive hazard in sodium cooled fast reactor, which is probably one of the main contributors to the total reactor risks. In this paper, the methodology of fast reactor sodium fire risk assessment was studied, following the introduction of the sodium fire. The application of this technology in China Experimental Fast Reactor was explored, and the results show that the core damage frequency induced by the sodium fire in reactor hall is 1.19 × 10-8(reactor · year). After that, several key problems which need to be further researched in the future during the process of sodium fire probabilistic safety assessment were discussed. (authors)

  15. Experimental study of the neutronics of the first gas cooled fast reactor benchmark assembly (GCFR phase I assembly)

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyya, S.K.

    1976-12-01

    The Gas Cooled Fast Reactor (GCFR) Phase I Assembly is the first in a series of ZPR-9 critical assemblies designed to provide a reference set of reactor physics measurements in support of the 300 MW(e) GCFR Demonstration Plant designed by General Atomic Company. The Phase I Assembly was the first complete mockup of a GCFR core ever built. A set of basic reactor physics measurements were performed in the assembly to characterize the neutronics of the assembly and assess the impact of the neutron streaming on the various integral parameters. The analysis of the experiments was carried out using ENDF/B-IV based data and two-dimensional diffusion theory methods. The Benoist method of using directional diffusion coefficients was used to treat the anisotropic effects of neutron streaming within the framework of diffusion theory. Calculated predictions of most integral parameters in the GCFR showed the same kinds of agreements with experiment as in earlier LMFBR assemblies.

  16. IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Presentations

    International Nuclear Information System (INIS)

    The fast reactor, which can generate electricity and breed additional fissile material for future fuel stocks, is a resource that will be needed when economic uranium supplies for the thermal reactors diminish. Further, the fast-fission fuel cycle in which material is recycled (a basic requirement to meet sustainability criteria) offers the flexibility needed to contribute decisively towards solving the problem of growing “spent” fuel inventories by greatly reducing the volume, the heat load and the radiotoxic inventory of high-level wastes that must be disposed of in long-term geological repositories. This is a waste management option that will play an increasingly important role in the future, and help to ensure that nuclear energy remains a sustainable long-term option in the world’s overall energy mix. In recognition of the fast reactor’s importance for the sustainability of the nuclear option, currently there is worldwide renewed interest in fast reactor technology development, as indicated, e.g., by the outcome of the Generation IV International Forum (GIF) technology review, which concluded with 3 out of 6 innovative systems to be fast reactors (gas cooled fast reactor, sodium cooled fast reactor, and heavy liquid metal cooled fast reactor), plus a potential fast core for a 4th concept, the super-critical water reactor. Currently, fast reactor construction projects are ongoing in India (PFBR) and Russian Federation (BN-800), whilst in China the first experimental fast reactor (CEFR) is in the commissioning phase. Fast reactor programs are also carried out in Europe (in particular in France), Japan, Republic of Korea and the USA. The most important challenges for fast reactors are in the areas of cost competitiveness with respect to LWRs and other energy sources, enhanced safety, non-proliferation, and public acceptance. With the exception of this latter, these translate into technology development challenges, i.e. the development of advanced reactor

  17. Measurements of power profile of the BN-600 commercial fast reactor by gamma-scanning and analytical studies of experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Izotov, V. V.; Kotchetkov, A. L.; Moiseev, A. V.; Semyonov, M. Y.; Seryogin, A. S.; Prishchepa, V. V.; Khomyakov, Y. S.; Tsyboulya, A. M. [State Scientific Center of the Russian Federation, Inst. for Physics and Power Engineering, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation); Dubrovsky, V. V.; Zheltyshev, V. A.; Ivanov, A. A.; Lyzhin, A. A.; Maltsev, V. V.; Mitrofanov, S. Y.; Roslyakov, V. F. [Branch of Rosenergoatom Concern, Beloyarsk NPP, Zarechny Sverdlovsk Region 624250 (Russian Federation); Belov, A. A.; Pryanitchnikov, A. V.; Seleznyov, E. F. [All-Russian Research Inst. on Operation of Nuclear Power Plants (VNIIAES), 25 Ferganskaya, Moscow 109507 (Russian Federation); Vasiliev, B. A.; Farakshin, M. P. [Experimental Design Bureau of Mechanical Engineering, 15 Burnakovsky Proezd, Nizhny Novgorod 603074 (Russian Federation)

    2006-07-01

    During 25 years of operation of BN-600 fast reactor at the Beloyarsk NPP, complex of analytical and experimental measurements has been developed for control of power rate distribution in the reactor core. Continuous control is performed by computational accompaniment based on three-dimensional multi-group analysis in hexagonal geometry in diffusion approximation. Periodical control is made by measuring of power rate profile in the standard fuel subassemblies of the BN-600 reactor by gamma scanning method on the stages of updating of the reactor core. By now, two cycles of such measurements have been performed when changing for the new reactor core design 01M2 providing 4-fold refueling mode and max fuel burn-up increased up to {approx}11.1% h.a. In the paper given are brief description of analytical and experimental methods of monitoring of power profile of the BN-600 reactor, results of their comparison and estimation of their precision based on the results of the above studies. It has been demonstrated that the use of 26-group diffusion approximation and GEFEST, JARFR and TRIGEX codes with ABBN-93 nuclear data gives adequate description of power rate distribution among the SAs of the BN-600 reactor core. Conservative estimation of calculation error is 5%. The main concern is evaluation of power profile of peripheral areas of the radial blanket and in-vessel storage, if achieved accuracy of 10-15% is insufficient. (authors)

  18. On-site experimental dynamic analysis for evaluating the soil-structure interaction and the seismic behaviour of the Italian PEC fast reactor building

    International Nuclear Information System (INIS)

    The paper describes the on-site dynamic tests carried out on the PEC fast reactor building, using various excitation methods (two eccentric back-rotating-mass mechanical vibrator, blasting in bore-hole, hydraulic actuators at the building foundations). It points out the purposes of the four tests campaigns performed at various construction stages and reports the main experimental results. These results concern both the design safety margins and the data for the validation of the three-dimensional numerical model of the reactor building, including soil-structure interaction phenomena. (author)

  19. Fast neutron flux in heavy water reactors

    International Nuclear Information System (INIS)

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author)

  20. Evaluation on Calculation Accuracy of the Sodium Void Reactivity for Low Void Effect Fast Reactor Cores with Experimental Analyses

    International Nuclear Information System (INIS)

    Calculation accuracy of the sodium void reactivity for safety-enhanced fast reactor core concepts was evaluated with analyses of critical experiments. In these concepts, heterogeneous core configuration and sodium plenum replacement are adopted to reduce the sodium void reactivity to around zero. In the past, a variety of critical experiments for heterogeneous cores had been carried out in the ZPPR facility, some of which are compiled in the IRPhEP handbook. Further, several experiments for core with sodium plenum had been performed in the BFS-2 facility. Calculation analyses of above mentioned critical experiments have been performed by using the Japanese current reactor physics analytical system. These analyses clarified that accuracy for homogeneous and axially-heterogeneous cores was sufficient, though accuracy for the radially-heterogeneous core and/or core with sodium plenum was not satisfactory. In order to achieve satisfactory accuracy for various types of cores, investigation on several design methods was performed. (author)

  1. Fast reactor fuel cycle facility

    International Nuclear Information System (INIS)

    An integrated fuel cycle facility named Fast Reactor Fuel Cycle Facility (FRFCF) is planned to be set up at Kalpakkam to close the fuel cycle of the Prototype Fast Breeder Reactor (PFBR) that is already under construction there. FRFCF is the first project of its kind in India. Closure of fuel cycle of PFBR will be a significant milestone of the second stage of nuclear power programme of the Department of Atomic Energy. The facility would be ready for operation in 2014. Design work and safety review of FRFCF are presently in progress. (author)

  2. Fast reactor designs: Commercial size fast reactors (unforeseen events)

    International Nuclear Information System (INIS)

    This chapter contains detailed design data and main operational data on the following commercial fast reactors (unforeseen events): Super-Phenix-1; Super-Phenix-2; SNR-2; BN-800; DFBR; CDFR; EFR; BN-1600; BN-1800; BREST-1200; JSFR-1500

  3. 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

  4. A review of the UK fast reactor programme, March 1979

    International Nuclear Information System (INIS)

    The Status report of the UK activities related to fast-breeder reactor activities includes the following: summary of the operating experience of the prototype Fast Reactor (PFR) during 1978; design studies of the commercial demonstration fast reactor (CDFR); design studies of later advanced LMFBR; engineering developments of high temperature sodium loop, steam generators and instrumentation; materials development; corrosion problems; sodium technology; fuel elements development; PFR fuel reprocessing; safety issues molten fuel-coolant interaction; core structure test; accident analysis; reactor performance studies; experimental reactor physics; fuel management and general neutronics calculation for CDFR; reactor instruments

  5. Experimental investigations on carbonation of sodium aerosol generated from sodium fire in the context of fast reactor safety

    International Nuclear Information System (INIS)

    Highlights: • Investigated carbonation of aerosol released from sodium fire. • Investigated influence of % RH and CO2 content in air on carbonation process. • Deployed a novel technique for assay of different specimens in Na-aerosol. • This work can be used to predict chemical assay of aerosols in large sodium fire. - Abstract: Carbonation of sodium aerosols is the most important aspects to be considered for the evaluation of chemical hazards as a part of fast reactor safety studies. The sodium oxide, immediately formed as the combustion product due to sodium fire, undergoes chemical changes to NaOH, Na2CO3 and NaHCO3 upon reactions with moisture and CO2 prevailed in the atmosphere. Of which, hydroxide aerosols are highly corrosive and harmful, and it has stringent concentration limit for human exposure. Hence, in order to assess the condition for human intervention in the event of sodium fire, chemical composition of aerosols resulting from controlled sodium fires in a closed Aerosol Test Facility was investigated. The real time chemical species of aerosols generated from sodium fire and the effect of relative humidity (RH) and carbon dioxide concentration in air on carbonation have been studied. The experiments were carried out with the initial mass concentration of ∼4 g m−3, RH between 20% and 90% and the CO2 concentration in surrounding environment at 390 and 280 ppm. It is observed from the experimental study that aerosols are enriched with NaOH (0.8 mol fraction) in the beginning stage (samples collected during first few minutes after sodium fire) when surrounding atmosphere contains any of the following compositions – (i) ∼90% RH and 390 ppm CO2, (ii) ∼90% RH and 280 ppm CO2 or (iii) 50% RH and 280 ppm CO2 whereas they are almost equally distributed between NaOH and Na2CO3 in the beginning stage when the atmosphere has any of the compositions (i) 50% RH, 390 ppm CO2, (ii) 20% RH, 390 ppm CO2 or (iii) 20% RH, 280 ppm CO2. Carbonation of

  6. Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo. Development of repair techniques for UCS replacement of Joyo

    International Nuclear Information System (INIS)

    Development of repair techniques in the reactor vessel of sodium cooled fast reactors is important to secure its safety and integrity. With the incident as an opportunity, repair techniques for Upper Core Structure (UCS) replacement was developed in Joyo. Since the UCS of Joyo was designed as an eternal structure and it has high dose rate due to the irradiation for over 30 years, the following subjects were mainly discussed in this study as critical tasks. (1) Prevention of deformation during jack-up and retrieval of the UCS, (2) Reducing UCS cask weight In order to resolve above (1), the allowable load to prevent the deformation of the UCS and guide sleeve were evaluated using finite element method analysis code with 3-D model. Furthermore, the general design of the equipment were established based on the requirement of monitoring and control of the factors (load, flatness, inclination and pull-up speed etc.) and gap observation to achieve UCS jack-up and retrieval. Concerning with above (2), multistage cask without crane and door valve was suggested in this study. As a result of this study, total weight of cask (including UCS) was reduced to be less than 100ton, which is the maximum load of the crane in the reactor containment vessel of Joyo. The UCS replacement is scheduled in 2014. Achievement of the UCS replacement and accumulated experience will be able to provide valuable insights for further improving and verifying repair techniques in sodium cooled fast reactors. (author)

  7. Experimental study on pressure loss in coated-particle-type fuel assembly of gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Revealing thermal-hydraulic characteristics in coated-particle fuel and fuel compartment regions of a fuel assembly is necessary for a design study of helium-gas-cooled fast reactors which is being performed at JNC as part of the feasibility studies on the commercialized fast reactor cycle system. This report describes a series of experiment to obtain basic data of pressure loss in particle layers and several porous materials simulating those of the fuel assembly as the first step of the investigation. A pressure loss measurement system was designed and manufactured in which the working fluid was dried nitrogen gas. Packed particle beds consisting of glass particles or lead ones were set to the test section of the system to get the basic pressure loss data of the coated-particle fuel region. Sintered metallic (bronze), laminated and perforated panels, and fibrous materials (wood, felt) were selected as the candidate structures of the fuel compartment region. These measurement were carried out under the several pressure levels (1 MPa-7 MPa). From the result of the measurement experiment, it was confirmed that Ergun's correlation equation related to packed beds gives appropriate pressure loss estimation of the coated-particle fuel region under the rated pressure condition and depressurized one and that modified Ergun's equation is applicable to the pressure loss of the sintered metallic. (author)

  8. Fast reactor research in Switzerland

    International Nuclear Information System (INIS)

    The small Swiss research program on fast reactors serves to further understanding of the role of LMFR for energy production and to convert radioactive waste to more environmentally benign forms. These activities are on the one hand the contribution to the comparison of advanced nuclear systems and bring on the other to our physical and engineers understanding. (author)

  9. Fast reactor savants take stock

    International Nuclear Information System (INIS)

    Some of the argument grew almost fierce when the Royal Society, one of the world's premier scientific institutions, held an international meeting in London last May on the fast neutron breeder reactor. Discussion skipped between broad scientific principles, technical minutiae, economics and politics. Some impressions are given by an independent writer on energy affairs. (author)

  10. Fast reactor savants take stock

    Energy Technology Data Exchange (ETDEWEB)

    Conway, Arthur

    1989-08-01

    Some of the argument grew almost fierce when the Royal Society, one of the world's premier scientific institutions, held an international meeting in London last May on the fast neutron breeder reactor. Discussion skipped between broad scientific principles, technical minutiae, economics and politics. Some impressions are given by an independent writer on energy affairs. (author).

  11. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  12. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  13. Status of national programmes on fast reactors

    International Nuclear Information System (INIS)

    Based on the International Working Group on Fast reactors (IWGFR) members' request, the IAEA organized a special meeting on Fast Reactor Development and the Role of the IAEA in May 1993. The purpose of the meeting was to review and discuss the status and recent development, to present major changes in fast reactor programmes and to recommend future activities on fast reactors. The IWGFR took note that in some Member States large prototypes have been built or are under construction. However, some countries, due to their current budget constraints, have reduced the level of funding for research and development programmes on fast reactors. The IWGFR noted that in this situation the international exchange of information and cooperation on the development of fast reactors is highly desirable and stressed the importance of the IAEA's programme on fast reactors. These proceedings contain important and useful information on national programmes and new developments in sodium cooled fast reactors in Member States. Refs, figs and tabs

  14. Status of fast reactor activities in Brazil

    International Nuclear Information System (INIS)

    This text describes the present status of fast reactor activities in Brazil, emphasizing the strategies being used to preserve this reactor concept as a viable alternative for future electricity generation in the country. The program is mostly research-oriented and has the objective of establishing a consistent knowledge basis which can serve as a support for the transition to the activities more directly related to design, construction and operation of an experimental fast reactor. Due to the present economic difficulties, the program is still modest but it is gradually growing. A report which has been finalized in December, 1995 and submitted to the authorities indicates the existence of the grounds for enlarging and consolidating the program. (author)

  15. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants

  16. Experimental study on reactivity of structural concrete with sodium-hydroxide in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    For countermeasure against sodium leak, structural concrete is protected by steel liner in a sodium-cooled fast reactor (SFR). However, if considering severe and unexpected accidental condition such as breach of steel liner by intensive sodium leak, the reaction of concrete with liquid sodium potentially may occur. For the purpose of elucidating the mechanism of sodium-concrete reaction in SFR, kinetic study of the sodium-hydroxide (NaOH)-silica (SiO2) reaction was carried out by Differential Scanning Calorimetry (DSC). The parameters, including melting point of NaOH, phase transition temperature of NaOH and SiO2, and NaOH-SiO2 reaction temperature were identified from DSC curves. From visualization test, sample eruption was observed during reaction. It was found that rate of NaOH-SiO2 reaction was quite fast from DSC curves, which was similar with that of the reaction between NaOH and aggregate of practical concrete. Thermal analysis results indicated that NaOH-SiO2 reaction could occur in the timeframe of sodium-concrete reaction. (author)

  17. Simulating the Behaviour of the Fast Reactor Joyo (Draft)

    International Nuclear Information System (INIS)

    Motivated by the development of fast reactors the behaviour of the Japanese experimental fast reactor Joyo is simulated with two Monte Carlo codes: Monte Carlo NParticle (MCNP) and Probabilistic Scattering Game (PSG). The simulations are based on the benchmark study 'Japan's Experimental Fast Reactor Joyo MKI core: Sodium-Cooled Uranium-Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO2 Blanket'. The study is focused on the criticality of the reactor, control rod worth, sodium void reactivity and isothermal temperature coefficient of the reactor. These features are calculated by applying both homogeneous and heterogeneous reactor core models that are built according to the benchmark instructions. The results of the two models obtained by the two codes are compared with each other and especially with the experimental results presented in the benchmark. (author)

  18. Advanced simulation for fast reactor design

    International Nuclear Information System (INIS)

    Full text: This talk broadly reviews recent research aimed at applying advanced simulation techniques specifically to fast neutron reactors. By advanced simulation we generally refer to attempts to do more science-based simulation - that is, to numerically solve the three-dimensional governing physical equations on fine scales and observe and study the holistic phenomena that emerge. In this way simulation is treated more akin to a traditional physical experiment, and can can be used both separately and in conjunction with physical experiments to develop more accurate predictive theories on reactor behavior. Many existing fast reactor modeling tools were developed for last generation's computational resources. They were built by engineers and physicists with deep physical insight - insight that both shaped and was informed by existing theory, and was underpinned by a vast repository of experimental data. Their general approach was to develop models that were tailored to varying degrees to the details of the reactor design, using free model parameters that were subsequently calibrated to match existing experimental data. The resulting codes were thus extremely useful for their specific purpose but highly limited in their predictive capability (neutronics to a lesser degree). They tended to represent more the state-of-the-art in our understanding rather than tools of exploration and innovation. Recently, a number of researchers have attempted to study the feasibility of solving more fundamental governing equations on realistic, three-dimensional geometries for different fast reactor sub-domains. This includes solving the Navier-Stokes equations for single-phase sodium flow (Direct Numerical Simulation, Large Eddie Simulation, and Reynolds Averaged Navier Stokes Equations) in the core, upper plenum, primary and intermediate loop, etc.; the non-homogenized transport equations at very fine group, angle, and energy discretization, and thermo-mechanical feedback based on

  19. 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

  20. Fuel management codes for fast reactors

    International Nuclear Information System (INIS)

    The CAPHE code is used for managing and following up fuel subassemblies in the Phenix fast neutron reactor; the principal experimental results obtained since this reactor was commissioned are analyzed with this code. They are mainly concerned with following up fuel subassembly powers and core reactivity variations observed up to the beginning of the fifth Phenix working cycle (3/75). Characteristics of Phenix irradiated fuel subassemblies calculated by the CAPHE code are detailed as at April 1, 1975 (burn-up steel damage)

  1. International Experience with Fast Reactor Operation and Maintenance

    International Nuclear Information System (INIS)

    This paper reviews the most important lessons learned from operation of the world’s sodium cooled fast reactors, both test reactors and power producing reactors, which represent nearly 400 reactor-years of cumulative operating experience. The first reactor in the world to produce electricity was a fast reactor, the Experimental Breeder Reactor I, in December 1951. International experience with fast reactor technology exists in France, Germany, India, Japan, the Russian Federation, the United Kingdom and the United States of America. The operating experience with these reactors has been mixed; early problems were associated with fuel cladding, steam generators, fuel handling and sodium leakage. Excellent experience has been gained, however, that demonstrates the robust nature of the technology, the potential for exceedingly safe designs, ease of maintenance, ease of operation and the ability to effectively manage waste from spent fuel. It is a mature technology. (author)

  2. Review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    This report covers the activities of the experimental fast reactor JOYO from April 1985 to March 1986. After completion of the 7th duty cycle operation at the end of March 1985, special operation was carried out for the in-vessel performance test of the failed fuel detection and location system by irradiating slitted pins, natural circulation test from 30 MWt, and in-core measurement of coolant flow rate of each core subassembly during April 1985

  3. 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,

  4. Analysis of Cm contained in irradiated fuel of experimental fast reactor JOYO'. Development of the analytical technique and measurement of Cm

    International Nuclear Information System (INIS)

    The analytical technique for Cm contained in a MOX fuel was developed and analysis of Cm contained in irradiated fuel of experimental fast reactor JOYO' was carried out, to contribute to evaluation of transmutation characteristics of MA nuclide in the fast reactor. The procedure of ion-exchange separation of Cm with nitric acid-methanol mixed media essential for the isotopic analysis in irradiated MOX fuel was adopted considering for being rapid and easy. The fundamental test to grasp separation characteristics of this procedure, such as Cm elution position and separation capacity between Cm and Am or Eu, was carried out. In applying this procedure to the analysis of Cm contained in actual specimen, separation condition was evaluated and optimized, and the procedure consist of impurity removal and Am removal process was devised. This procedure resulted in high recovery rate of Cm and high removal rate of Am and impurity which becomes a problem in sample handling and mass-spectrometry such as Eu and Cs. The Cm separation test from irradiated MOX fuel was carried out using this technique, and Cm isotopic ratio analysis was enabled. The analytical technique for Cm contained in irradiated MOX fuel was established using the procedure of ion-exchange separation with nitric acid-methanol mixed media. The analysis of Cm contained in irradiated MOX fuel of experimental fast reactor 'Joyo' was carried out. As a result, it was revealed from measured data that Cm content rate was 1.4-4.0x10-3 atom%, small amount of 247Cm was generated and Cm isotopic ratio was constant above burn-up 60 GWd/t. (author)

  5. Experimental verification of fast reactor safety analysis code SIMMER-III for transient bubble behavior with condensation

    International Nuclear Information System (INIS)

    Experimental verification of a reactor safety analysis code, SIMMER-III, was performed for transient behaviors of large-scale bubbles with condensation. The objective of the present study is to verify the code for numerical simulations of relatively short-time-scale multi-phase, multi-component hydraulic problems, among which vaporization and condensation, or simultaneous heat and mass transfer, play an important role. In this study, a series of transient bubble behavior experiments, which are dedicated to condensation phenomena with noncondensable gases, was performed. In the experiments, pressurized mixture of noncondensable gas and steam was discharged as a large-scale single bubble into a cylindrical pool filled with stagnant subcooled water. Concentration of noncondensable gas was taken as an experimental parameter as well as species of noncondensable gas. The characteristics of transient behavior of large-scale bubbles with condensation observed in the experiments were estimated through experimental analyses using SIMMER-III. In the experiments with steam condensation, dispersion of the gas mixture discharged into the liquid pool was accompanied by the vapor condensation at the bubble surface. SIMMER-III simulations suggest that the noncondensable gas has less inhibiting effect on condensation of large-scale bubbles. This is a characteristic different from quasi-steady condensation of small-scale bubbles observed in our previous experiments. (author)

  6. Fatigue experiment on bursting disk device in non-active heat discharging system of chinese experimentation fast reactor

    International Nuclear Information System (INIS)

    Passive residual heat removal system is very important for faster reactors to ensure its safety. Bursting disk is one passive discharging device under over-pressure. Action performance of the bursting diak is carried out by adding fatigue loads repeatedly in this paper. The load stress is 80% of the nominal bursting pressure of the bursting disk, and the load cycles is over 105. The experimental result shows that the action pressure difference before and after the fatigue experiment of the bursting disk is less than 1.5%, which indicates that the fatigue resistance of the bursting disk for the passive residual heat removal system is excellent, and it can ensure the reliable work of the passive system. (authors)

  7. Heterogeneous Recycling in Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  8. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  9. Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity - Irradiation Tests of Np and Np-U Samples in Experimental Fast Reactor Joyo (JAEA) and Advanced Thermal Reactor ATR (INL)

    International Nuclear Information System (INIS)

    A project of Produce Protected Plutonium (P3) was proposed by Tokyo-Tech as a part of non-proliferation research for Plutonium (Pu) utilization to nuclear reactor. The project is to reach the production of inherently protected Pu by addition 237Np to Uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. Experimental determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor Joyo was done to evaluate 238Pu production from 237Np under the fast neutron spectra. The amount of 238Pu in the irradiated 237Np samples was determined by a radiochemical analysis in Alpha-gamma Facility of JAEA. The produced 238Pu in the samples was found to depend on the neutron spectrum, ranging from that of a typical fast reactor to a nearly epi-thermal spectrum. The fast reactor can potentially control the 238Pu production from 237Np by the spectrum shift in the different irradiation position. Within the framework of the P3, the fast reactor can make roles of protected Pu production which can be better performed in the reflector region, the ratio of 238Pu is achieved up to around 90% produced from 237Np. On the other hand, 2, 5 and 10% 237Np containing U samples were also irradiated in Advanced Thermal Reactor of INL to evaluate the 238Pu production under thermal neutron region. The irradiation condition and its loading position of samples were fixed based on a calculation with MCWO (MCNP Coupled with ORIGEN2) code. The fuel specimens were removed from the core at 100, 200 and 300 effective full power days (EFPD), and then post irradiation examination was completed at Chemical lab. in MFC of INL. For the samples after 300 EFPD irradiation, Np depletion were about 60 % for 2% Np-U samples, about 50% for 5 and 10 % Np-U samples. The 238Pu to Pu ratio was about 20%, 30% and 45% for 2%, 5%, and 10% Np-U samples, respectively. The neutronics calculation results were coincident with the experimental ones. Acknowledgments: The

  10. 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

  11. Spatial Kinetics in Fast Reactors

    International Nuclear Information System (INIS)

    Reactor neutronic calculations designed for calculating of unsteady processes in a real 3D geometry require processing of a large amount of information. They cannot consist of simple models, as they should reflect the processes of variations of all local reactor characteristics. The model complexity and the significant time needed for numerical solution of neutron-transport equations limit the choice of methods that can achieve the required accuracy. Thus there is an urgent need for the development of various methods enabling the solution of unsteady neutron-transport equations and estimates of their errors, spent time and consistency with the experimental data. (author)

  12. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. Recovery of MARICO-2 sample part

    International Nuclear Information System (INIS)

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. The following items are introduced here: (1) summary of restoration work and overall process of restoration work, (2) recovery operation of MARICO-2 sample part, (3) exchange of the upper core structure that was conducted this year, and (4) results of recovery of MARIKO-2 sample part. (A.O.)

  13. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 1. MARICO-2 subassembly retrieval work

    International Nuclear Information System (INIS)

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. This paper introduces the progress of restoration work and the future work plan, with a focus on the outline of overall restoration work, the method / problems / measures for MARICO-2 sample part recovery operations, and fabrication of sample part recovery device. (A.O.)

  14. Experimental study for research and development of a super fast reactor. (2) Oscillatory condensation of high temperature vapor directly discharged into sub-cooled liquid pool

    International Nuclear Information System (INIS)

    The measurement of pressure oscillation and the observation of condensation behavior of a vapor discharge into sub-cooled liquid cool has been carried out to obtain a basic data for the evaluation of safety of the LOCA in the supercritical pressure light water cooled fast reactor (Super Fast Reactor). In the experiment, HCFC 123 is used as the test fluid. HCFC 123 is easy for handling due to its low critical pressure and temperature, and therefore, the experimental conditions can be set easily to make systematic data. The vapor at high temperature is discharged into the sub-cooled liquid pool through a submerged single pipe vertically fixed. The oscillatory condensation is observed. The condensation oscillation produces pressure oscillation in the liquid pool. The condensing interface area becomes small as the increase of the degree of sub-cooling. The pressure frequency has a period of millisecond order and the frequency and amplitude of the pressure oscillation increase with increasing the degree of sub-cooling and mass flux of the vapor, like the results of some conventional water vapor injection tests. In the present study, it is also consistently discussed the influence of the vapor temperature, mass flux, mass flow rate, back pressure of the liquid pool, pipe diameter and the degree of sub-cooling on the pressure amplitude and condensation behavior. (author)

  15. A review of the UK fast reactor programme. March 1977

    International Nuclear Information System (INIS)

    This paper reports on the Fast Reactor Programme of United Kingdom. These are the main lines: Dounreay Fast Reactor; Prototype Fast Reactor; Commercial Fast Reactor; engineering development; materials development; chemical engineering/sodium technology; fast reactor fuel; fuel cycle; safety; reactor performance study

  16. 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

  17. Current status of restoration work for obstacle and supper core structure in reactor vessel of experimental fast reactor 'JOYO'. 3. Sodium purification operation after MARICO recovery and UCS exchange work

    International Nuclear Information System (INIS)

    At fast-breeder reactor 'Joyo', in order to restore the partial inhibition of the rotating plug fuel exchange function due to interference with 'experimental apparatus with instrumentation lines (MARICO-2)', which occurred in May 2007, a recovery work was performed. The replacement work of the upper core structure and the recovery of sample part of the experimental apparatus with instrumentation lines were carried out under conditions where the primary system sodium was drained and the liquid level of reactor vessel was lowered. During the pulling-up work of upper core structure, an increase in nitrogen and hydrogen concentrations in the reactor vessel cover gas (argon) was confirmed through the measurement of the primary system gas chromatograph. This was due to the intrusion of air caused by the opening of the cover gas boundary. Since entrained oxygen reacted with sodium in the reactor, the purity of sodium was reduced. When this sodium is purified according to common method, the sodium with decreased purity defuses through the entire primary cooling system, causing various adverse effects. A safe and reliable procedure to purify sodium while preventing the adverse effects was examined and practiced. (A.O.)

  18. Research activities on fast reactors in Switzerland

    International Nuclear Information System (INIS)

    The current domestic Swiss electricity supply is primarily based on hydro power (approximately 61%) and nuclear power (about 37%). The contribution of fossil systems is, consequently, minimal (the remaining 2%). In addition, long-term (but limited in time) contracts exist, securing imports of electricity of nuclear origin from France. During the last two years, the electricity consumption has been almost stagnant, although the 80s recorded an average annual increase rate of 2.7%. The future development of the electricity demand is a complex function of several factors with possibly competing effects, like increased efficiency of applications, changes in the industrial structure of the country, increase of population, further automation of industrial processes and services. Due to decommissioning of the currently operating nuclear power plants and expiration of long-term electricity import contracts there will eventually open a gap between the postulated electricity demand and the base supply. The assumed projected demand cases, high and low, as well as the secured yearly electric energy supply are shown. The physics aspects of plutonium burning fast reactor configurations are described including first results of the CIRANO experimental program. Swiss research related to residual heat removal in fast breeder reactors is presented. It consists of experimental ana analytic investigations on the mixing between two horizontal fluid layers of different velocities and temperatures. Development of suitable computer codes for mixing layer calculation are aimed to accurately predict the flow and temperature distribution in the pools. A satisfactory codes validation based on experimental data should be done

  19. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  20. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  1. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  2. BN800: The advanced sodium cooled fast reactor plant based on close fuel cycle

    International Nuclear Information System (INIS)

    As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc. (authors)

  3. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

    The fuel cycle for fast reactors, is analysed, regarding the technical aspects of the developing of the reprocessing stages and the fuel fabrication. The environmental impact of LMFBRs and the waste management of this cycle are studied. The economic aspects of the fuel cycle, are studied too. Some coments about the Brazilian fast reactors programs are done. (E.G.)

  4. Status of fast reactor research in Germany

    International Nuclear Information System (INIS)

    The paper gives a short survey of fast reactor activities in Germany. The fast reactor activities of FZK are part of the Nuclear Safety Projects. The R and D program include neutron physical and safety calculated, and post-irradiated examination of structural materials. The key issues and tasks of the program concerned safety and transmission of minor activities and fission products. (author)

  5. Startup operational tests of fast reactors

    International Nuclear Information System (INIS)

    This paper is mainly concerned with the experiences of the two main phases of startup operational tests of fast reactors: (1) The general tests and Sodium filling before core loading. (2) The core loading,approach to criticality and power build up operational tests, taking for example a large and middle demonstrating integrated-type fast reactor. (author)

  6. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  7. 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

  8. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  9. Fast reactor strategy in European Union

    International Nuclear Information System (INIS)

    The tendency and strategy of fast reactors development in European Union are considered. The advantages and disadvantages of sodium, lead-bismuth and gas cooled fast reactors are discussed. It is shown that development of such reactors is the further sustainable development of nuclear power engineering. All three tendencies have clear structure and tasks, all prototypes will appear by 2020 and NPP - towards the middle of the century. It is pointed out that sodium coolant is the leading tendency in fast reactor development in European Union

  10. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    • Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

  11. The 'RB' Reactor as a Source of Fast Neutrons

    International Nuclear Information System (INIS)

    A study of the RB reactor as possible source of fast neutrons began in 1976 and four different version of fast neutron sources are designed up to 1990: an external neutron converter - ENC (1976), an experimental fuel channel - EFC (1982), an internal neutron converter - INC (1983), and a coupled fast-thermal core - HERBE (1990). An overview of applications and characteristics of each particular source of fast neutrons, including available irradiation space, neutron spectra and equivalent neutron and gamma dose rates is presented in the paper. Control and safety-related implications of these modifications of the reactor are emphasised. Computer codes and nuclear data libraries, used in calculations, are described. (author)

  12. Fast reactor passive shutdown system: LIM

    International Nuclear Information System (INIS)

    To enhance the inherent safety of the fast breeder reactor (FBR), unique attempts are being made in reactivity control systems design to achieve maintenance-free and reliable performance. The design proposed is the lithium injection module (LIM) for inherent ultimate shutdown. Reactor physics calculation revealed the reactivity worth of LIM in a 60 MWe metal-fueled FBR and a 1,000 MWe mixed-oxide-fueled FBR. An experimental verification on the freeze seal design assured an accurate injection temperature of LIM. Reliability, maintainability, and real time monitoring for LIM is also discussed. A definite advantage over the conventional self-actuated shutdown system (SASS) has been presented. LIM offers substantial inherent safety with improved maintainability. (author)

  13. Integral physics data for fast-reactor design

    International Nuclear Information System (INIS)

    Integral physics data for fast-reactor design. The recent compilation of the section on fast-reactor physics for the forthcoming second edition of 'Reactor Physics Constants' has necessitated a survey of the available experimental integral data. The choice of fast-reactor-physics integral data to be included in the compilation was based upon two criteria besides availability: (a) the data arise from relatively simple systems which lend themselves to simple theoretical analyses; and (b) complicated systems representing prototypes or mock-ups having general interest in terms of fast-power reactors. The first criterion was decided upon so as to list integral data for those systems of most general utility for the verification of cross-section parameters and calculational procedures. The second criterion is based upon presentation of current data on actual fast power breeder reactor systems. These are too complicated for simple theoretical analysis. They demonstrate the complexity of the actual reactor versus the more idealized and easily analysed critical experiment. Integral physics data for reactor design refer to measurements on reactor systems, critical or otherwise, of the various reactor physics quantities of practical and/or theoretical importance. These characterize and lead to an understanding of the system. The measurements are represented by critical mass, core shape factor, detector ratios, neutron spectra, material replacement experiments, reflector savings, neutron lifetime, Rossi-α, and similar quantities. These data are reviewed and the range of applicability is described. Limitations of experimental and analytical results are shown to exist in certain spectral and criticality analyses. Experimental and analytical investigations are suggested for future work. These will tend to narrow the gap between theory and experiment on 'known' systems. They also include investigations to 'firm up' the physics of large conceptual, fast power-breeder reactor

  14. 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)

  15. ELSY - The European Lead Fast Reactor

    OpenAIRE

    Alemberti, Alessandro; Carlsson, Johan; Malambu, Edouard; ORDEN Alfredo; CINOTTI Luciano; STRUWE Dankward; Agostini, Pietro; Monti, Stefano

    2009-01-01

    The European Lead Fast Reactor is being developed starting from September 2006, in the frame of the ELSY (European Lead SYstem) project funded by the Sixth Framework Programme of EURATOM. The project, coordinated by Ansaldo Nucleare, involves a wide consortium of European organizations. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered te...

  16. Status of fast reactor activities in Russia

    International Nuclear Information System (INIS)

    This paper outlines state-of-the-art of the Russian nuclear power as of 1997 and its prospects for the nearest future. Results of the BR-10, BOR-60 and BN-600 reactors operation are described, as well as activity of the Russian institutions on scientific and technological support of the BN-350 reactor. Analysis of current status of the BN-800 reactor South-Urals NPP and Beloyarskaya NPP designs is given in brief, as well as prospects of their construction and possible ways of fast reactor technology improvement. Studies on fast reactors now under way in Russia are described. (author)

  17. Status of national programmes on fast reactors 1997/98. 31. annual meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    The objective of the meeting was to co-ordinate the exchange of information on the status of fast reactor development and operational experience, including experience with experimental types of reactor; to consider meeting arrangements for 1998 and 1999; and to review the IAEA co-ordinated research activities in the field of fast reactor, as well as co-ordination of the International Working Group on Fast Reactors activities with other organizations

  18. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  19. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  20. Some progress on seismic isolation technology in building structure in China

    International Nuclear Information System (INIS)

    Seismic isolation technology has been considerably developed in China. Appropriate codes and design manuals have ben used. There is a plan of developing Fast reactor technology in China. The conceptual design for a fast experimental reactor was completed. Investigation of seismic isolation technology for fast reactor has started

  1. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements

  2. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  3. An experimental study on heat transfer from a mixture of solid-fuel and liquid-steel during core disruptive accidents in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    The relocation of degraded core material through the Control Rod Guide Tubes (CRGTs) is one of essential subjects to achieve the in-vessel retention (IVR) in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The CRGT is available as the discharge path by its failure in the core region and heat-transfer from the core-material to the CRGT is one of dominant factors in its failure. In case of a core design into which a fuel subassembly with an inner duct structure (FAIDUS) is introduced, a mixture of solid-fuel and liquid-steel is supposed to remain in the core region since the FAIDUS could effectively eliminate fuel in liquid-state from the core region. Therefore, the objective of the present study is to obtain experimental knowledge for the evaluation of heat-transfer from the mixture of solid-fuel and liquid-steel to the CRGT. In the present study, an experiment was conducted using Impulse Graphite Reactor which is an experimental facility in National Nuclear Center of the Republic of Kazakhstan. In the experiment, the mixture of solid-fuel and liquid-steel was generated by a low-power nuclear heating of fuel and transferring its heat to steel, and then, data to consider the heat-transfer characteristics from the mixture of solid-fuel and liquid-steel to the CRGT were obtained. The heat-transfer characteristic was revealed by evaluating thermocouple responses observed in the experiment. Through the present study, knowledge was obtained to evaluate heat-transfer from the remaining core-materials to the CRGT. (author)

  4. Prospects for future reactor utilization in China

    International Nuclear Information System (INIS)

    The utilization of nuclear energy as district heating describes the situation of present residential heating, advantages of nuclear heat supply, adaptation of low-temperature nuclear heat reactors. It is planned first to build one or two low-temperature nuclear heat reactors in appropriate northern cities of China in the 1990s. Take the example of the city Qiqihar in Northeast China, it is estimated that if the district heat supply system installed with 400 MWth low temperature heat reactor, the consumption of around 300,000 tons coal could be saved, it would mitigate rail congestion and reduce environmental pollution. Utilization of nuclear energy as district heating is favorable in cold regions, that is the Northeast, Northwest and North China areas. Nuclear heating cost will be lower than conventional coal fire heating by comparison. The research work on FBR, the plan for building experimental FBR, and research work on fusion technology referring to advanced reactors are described. FBR, the next generation reactor, is important for economic utilization of nuclear fuel due to its inherent function of breeding fissile materials. Research work of FBR has been carried out for several years, 50-100 MWth experimental FBR will be built before the year 2000. Feasibility study of HTRGR has been carried out in recent years. Efforts have also been devoted to research and development work of the advanced light water reactors and fusion technology. China will vigorously promote the peaceful application of nuclear energy in national economy and its people's life by fully utilizing the existing nuclear industry and technology. International cooperation in nuclear field cannot only bring benefit to both developed and developing countries but accelerate the peaceful application of nuclear energy in the world

  5. Methods for quantifying uncertainty in fast reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H.; Fischer, P. F.

    2008-04-07

    Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

  6. Conceptual designs of advanced fast reactor. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    A Technical Committee meeting (TCM) was held on Conceptual Designs of Advanced Fast Power Reactors to review the lessons learned from the construction and operation of demonstration and near-commercial size plants. This TCM focused on design and development of advanced fast reactors and identified methodologies to evaluate the economic competitiveness and reliability of advanced projects. The Member States which participated in the TCM were at different stages of LMFR development. The Russian Federation, Japan and India had prototype and/or experimental LMFRs and continue with mature R and D programmes. China, the Republic of Korea and Brazil were at the beginning of LMFR development. Therefore the aims of the TCM were to obtain technical descriptions of different design approaches for experimental, prototype, demonstration and commercial LMFRs, and to describe the engineering judgements made in developing the design approaches. Refs, figs, tabs

  7. OECD Nuclear Energy Agency Activities Related to Fast Reactor Development

    International Nuclear Information System (INIS)

    The OECD Nuclear Energy Agency (NEA), whose role is to assist its member countries to develop, through international cooperation, the scientific and technological bases required for the safe, environmentally friendly and economical use of nuclear energy, conducts work related to fast reactor systems along two areas of activity: one focused on scientific research and technology development needs and one dedicated to strategic and policy issues. The paper summarizes recent and ongoing NEA activities in each of these areas of activity, including: improved evaluations of basic nuclear data needed for the development of fast reactor systems, expansion of integral experiments databases to provide improved validation for fast reactor modelling methods, modelling of transients in SFRs, creation of an innovative fuels expert group, a series of information exchange meetings on actinide and fission product partitioning and transmutation, study on homogeneous versus heterogeneous recycle of transuranic isotopes in fast reactors, studies on research needs and the availability of experimental facilities for fast reactor safety studies, and a study on trends towards sustainability in the nuclear fuel cycle. The NEA is also an active player in many other international activities related to fast neutron systems, such as the Generation-IV International Forum where the NEA acts as technical secretariat for the project. The NEA will continue to support member countries in the field of fast reactor development and related advanced fuel cycles by providing a forum for exchange of information and various other collaborative activities. (author)

  8. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    The fast breeder reactor development project in Japan made progress in the past year, and will be continued in the next fiscal 1981. The scale of efforts both in budget and personnel will be similar to those in fiscal 1980. The budget for R and D works and for the construction of the fast breeder prototype reactor ''Monju'' will be approximately 20 billion yen and 27 billion yen, respectively, excluding the wage of the personnel concerned. The number of the technical personnel currently engaging in fast breeder reactor development in the Power Reactor and Nuclear Fuel Development Corp. is about 530. As for the experimental fast reactor ''Joyo'', three operational cycles at 75 MWt have been completed in August, 1980, and the fourth cycle has started in March, 1981. As for the prototype reactor ''Monju'', progress was made toward the construction, and the environmental impact statement on the reactor was approved by the authorities concerned. The studies on the preliminary design of large LMFBRs have been made by the PNC and also by power companies. The design study carried out by the PNC is concerned with a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of the commissioning of ''Monju''. The highlights and topics in the development activities for fast breeder reactors in the past twelve months are summarized in this report. (Kako, I.)

  9. ELSY - The European lead fast reactor

    International Nuclear Information System (INIS)

    Heat Removal) systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; notwithstanding it has been demonstrated that the high density of lead can be mitigated by more compact solutions and improvement of the design of the Reactor Vessel support system, i.e. the adoption of seismic isolators for a full seismic-resistant design. Preliminary results of the reactor vessel and supports stress analysis indicate that an LFR larger than a medium-size plant (in the IAEA classification) is potentially feasible. Safety has been one of the major focus all over the ELSY development. In addition to the inherent safety advantages of lead coolant like high boiling point and no exothermic reactions with air or water, a high safety grade of the overall system has been reached. In fact overall primary system has been conceived in order to minimize pressure drops and, as a consequence, to allow decay heat removal by natural circulation (note that this feature is essential for the unprotected loss of flow transient). Moreover two redundant, diverse and passive operated DHR systems have been developed and adopted. A sketch of the ELSY primary system configuration is shown. The paper focuses on the main aspects of the proposed design for the European Lead Fast Reactor highlighting the innovation of this reactor concept, overall objectives as well as future developments. Main safety features of the proposed Decay Heat Removal systems will be presented. Some experimental results related to the development of appropriate materials or materials protection for the high corrosion environment of Lead are also briefly presented

  10. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  11. A Comparison of Theoretical and Experimental Values of the Activation Doppler Effect in Some Fast Reactor Spectra

    International Nuclear Information System (INIS)

    Results of activation Doppler measurements on the U238 (n,γ) and U235 (n, fission) reactions in the FR0 and MSCA fast critical assemblies have been compared with theoretical values. The study covers neutron spectra with median fission energies from 50 to 240 keV. The calculated Doppler effect in U238 in the FR0 cores is 20 - 35 % lower than the measured values. The sensitivity of the theoretical result with regard to changes in cross sections and neutron spectrum has been studied. The theoretical value for U235 (FR0 core 5) is 4 times higher than the measured one. The report includes a brief description of the DORIX-2 method of calculating effective resonance cross sections appropriate to activation Doppler measurements. References to the cross section data used for the comparisons are also given

  12. Experimental results from the BNL zero power reactor HITREX

    Energy Technology Data Exchange (ETDEWEB)

    Kitching, S.J.; Lewis, T.A.; Playle, T.S.

    1973-10-15

    This report presents experimental results obtained with the BNL reactor Hitrex. Measurements of reactivity, and of thermal and fast neutron reaction rate distributions have been made with various experimental control rod configurations.

  13. New fast-reactor approach

    International Nuclear Information System (INIS)

    The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel

  14. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

    Acknowledging the importance of passing on knowledge and experience, as well mentoring the next generation of scientists and engineers, and in response to expressed needs by Member States, the IAEA has undertaken concrete steps towards the implementation of a fast reactor data retrieval and knowledge preservation initiative. Decommissioning of fast reactors and other sodium bearing facilities is a domain in which considerable experience has been accumulated. Within the framework and drawing on the wide expertise of the Technical Working Group on Fast Reactors (TWG-FR), the IAEA has initiated activities aiming at preserving the feedback (lessons learned) from this experience and condensing those to technical recommendations on fast reactor design features that would ease their decommissioning. Following a recommendation by the TWG-FR, the IAEA had convened a topical Technical Meeting (TM) on 'Operational and Decommissioning Experience with Fast Reactors', hosted by CEA, Centre d'Etudes de Cadarache, France, from 11 to 15 March 2002 (IAEA-TECDOC- 1405). The participants in that TM exchanged detailed technical information on fast reactor operation and decommissioning experience with various sodium cooled fast reactors, and, in particular, reviewed the status of the various decommissioning programmes. The TM concluded that the decommissioning of fast reactors to reach safe enclosure presented no major difficulties, and that this had been accomplished mainly through judicious adaptation of processes and procedures implemented during the reactor operation phase, and the development of safe sodium waste treatment processes. However, the TM also concluded that, on the path to achieving total dismantling, challenges remain with regard to the decommissioning of components after sodium draining, and suggested that a follow-on TM be convened, that would provide a forum for in-depth scientific and technical exchange on this topic. This publication constitutes the Proceedings of

  15. Liquid Metal Coolant Technology for Fast Reactors

    International Nuclear Information System (INIS)

    In the paper presented are results of comparative analysis and the choice of liquid metal coolants for fast reactors, the current status of studies on the physical chemistry and technology of sodium coolants for fast neutron reactors and heavy liquid metal coolants, namely, lead-bismuth and lead for fast reactors and accelerator driven systems. There are descriptions of devices designed for control of the impurities in sodium coolants and their removal as well as methods of heavy liquid metal coolant quality control, removal of impurities from heavy liquid metal coolants and the steel surface of components of nuclear power plants (NPPs) and relevant equipment. Attention is given to the issues of modelling of impurity mass transfer in liquid metal coolants and designing new liquid metal coolants for NPPs. Results of the analysis of NPP abnormal operating conditions are presented. The adopted design approaches assure reliable protection against accidents. Up to now, about 200 reactor-years of sodium cooled fast reactor operation and about 80 reactor-years of submarine reactor operation have been gained. The new goals for sodium and heavy liquid metal coolant technology have been formulated as applied to the new generation fast reactors. (author)

  16. Statement to International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, 7 December 2009, Kyoto, Japan

    International Nuclear Information System (INIS)

    Full text: Distinguished Guests, Ladies and Gentlemen, It is my honour to address participants at this opening session of the International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, organized by the IAEA and hosted by the Japan Atomic Energy Agency. Fast reactor technology has the potential to ensure that energy resources which would last hundreds of years with the technology we are using today will actually last several thousand years. In other words it can withstand enormous increases in demand. This innovative technology also reduces the risk to the environment and helps to limit the burden that will be placed on future generations in the form of waste products. The coming year will be an exciting one for the development of fast-spectrum nuclear reactors. We expect to reach many important milestones: - the first criticality of the China Experimental Fast Reactor; - the restart of the Monju prototype fast reactor in Japan; and - the new insights we will gain through the end-of-life studies at the Phenix reactor in France. In the near future, new fast reactors will be commissioned: the 500MW(e) Prototype Fast Breeder Reactor in India, the first in a series of five of the same type, and the BN-800 reactor in the Russian Federation. Moreover, France, Japan, India, China and the Republic of Korea are preparing advanced prototypes, demonstration or commercial reactors for the 2020-2030 period. Nuclear power is set to be an increasingly important part of the global energy mix in the coming decades as demand for energy grows. Scores of countries in both the developed and developing world have told the IAEA that they are interested in introducing nuclear power. The 30 countries which already have nuclear power reactors are set to build more. This trend is likely to be accompanied by accelerated deployment of fast reactors. Continued advances in research and technology development are necessary to ensure improved economics and

  17. Fast neutron benchmark proposal at TRIGA-ACPR Reactor

    International Nuclear Information System (INIS)

    The development of fast neutron benchmarks is a historical aim of reactor physics. The dry experimental tube situated in the central region of the core in TRIGA Annular-Core Pulsing Reactor (ACPR) offers a suitable neutron source for fast neutron benchmark development. Our proposal consists in mounting a high-enriched uranium annular converter into the dry channel of the core. Preliminary computations and measurements are presented in this paper. Neutron flux computations in the dry channel and the uranium converter were performed using MCNP and WIMS codes. Also neutron flux spectrum measurements and fast and thermal neutron flux distribution measurements were performed using foil activation techniques. (authors)

  18. Review of the analyses of the Doppler-effect measurements in SEFOR [Southwest Experimental Fast Oxide Reactor

    International Nuclear Information System (INIS)

    The SEFOR experimental results and the three original analyses are reviewed and discussed. The emphasis of the review is placed on aspects that are pertinent to a possible modern re-analysis of the experimental results. Looking at the analysis results in terms of zero and first order effects shows that the zero order effects, the Doppler constant of the two SEFOR cores, are obtained by the three analyses in satisfactory agreement. But the first order effects, but temperature variation of this Doppler-constant quantity, cannot be determined with any informative accuracy. Since this is likely due to limitations in the experiments, a re-analysis - except for methodological reasons - does not appear to be fruitful. 17 refs., 8 figs., 7 tabs

  19. Evaluation on Calculation Accuracy of the Sodium Void Reactivity for Low Void Effect Fast Reactor Cores with Experimental Analyses

    International Nuclear Information System (INIS)

    Conclusion: • Larger calculation uncertainty of sodium void reactivity in low or negative void effect core concepts and requirement of much finer treatment for these cores in comparison with conventional homogeneous core concepts; • Some or considerable impact to transient analysis by reflecting information of the FBR core design database; • Importance of extension on sodium void reactivity experimental data and their reflection to enhance the reliability of safety analysis

  20. Experimental investigations of local flow parameters near the walls of fuel elements of fast breeder reactors. Pt. 2

    International Nuclear Information System (INIS)

    The here presented material is the result of the first stage of experiments performed on the NEM-2 model, with the initial comparative values for evaluation of the effects of the geometry dimensions in the cluster upon the fluid-dynamic conditions being concerned in the main. The investigation of those effects (presence of displacement bodies, displacement of single rods or rod groups etc.) will be subject of further experimental works. (orig.)

  1. Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  2. Fast Reactor Physics Vol. I. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  3. Current status on fast reactor program in Kazakhstan

    International Nuclear Information System (INIS)

    Atomic scientific-industrial complex of Republic of Kazakhstan consist of: Uranium mining, production and power industry, Enterprises of uranium ores geological searching and number of natural mines (using the mining and underground leaching techniques); Two plants of U3O8 production at Aktau and Stepnogorsk towns; Metallurgical plant producing uranium fuel pellets for fuel assemblies of RBMK and VVER reactors types; Energy plant at Aktau (MAEK) is used for production of heat, electricity and desalination of water and based on three energy blocks using natural gas and one nuclear unit with fast breeder reactor BN-350. The fast breeder reactor BN-350 at Aktau was commissioned in November 1972 and finally shutdown in April 1999. Three different type of the research reactors and non reactor test facility on the territory of the former Semipalatinsk Nuclear Test Site and one research reactor and subcritical assembly nearly Almaty are exploiting for the investigation in field of reactors nuclear safety and other type of investigations. These are: VVR-K - light water reactor, power - 10 MW, EWG-1M - thermal light water heterogeneous vessel reactor with light water moderator and coolant, beryllium reflector, maximum thermal power - 35 MW, IGR - impulse homogeneous uranium-graphite thermal neutron reactor with graphite reflector, RA - thermal neutron high temperature gas heterogeneous reactor with air coolant, zirconium hydride moderator, and beryllium reflector, about 0.5 MW power, EAGLE - non reactor test facility for reactor fuel element melt process due to severe accident studding. Project on construction of experimental reactor TOKOMAK at city Kurchatov (in frame of International Thermonuclear Experimental Reactor) is going on (design and equipment manufacture and procurement stage). Accomplishment of the project is estimated for year 2007. Works on construction of the new cyclotron at Astana University started at the beginning of this year in co-operation with Dubna

  4. Experimental and theoretical study of a particular type of transient flow of boiling sodium: flow excursion. Study carried out in the framework of the fast neutron reactor safety

    International Nuclear Information System (INIS)

    The flow excursion phenomenon was studied in the framework of safety analyses for sodium cooled fast reactors, as a consequence of an accident of pump breakdown without safety rod drop. The experimental study, performed on an out-of-pile circuit with sodium forced convection allowed the flow excursion due to boiling appearance in a single heating channel feeded in parallel with a by-pass to be investigated. The first phase is characterized by a relatively slow decay of the (mean) flow rate through the channel from its initial value (corresponding to starting boiling) up to a mean value near zero; its duration, rather important, is of the order of ten or thirty seconds. The second phase appears when the boiling zone occupies an important fraction (half or three quarters) of the heating length of the channel. The mean flow rate has a low value, near zero; the flow rate is oscillatory (chugging) and is formed of vapor plugs parted by liquid bundles. This chugging phase lasts two or three seconds. The computation model developed described simply the boiling liquid flow (homogeneous sliding model) and the heat transfer between the heating medium and the fluid. The transient terms are neglected in continuity and energy equations so as the fluid in the channel is incompressible and its thermal inertia nul

  5. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties

  6. Development in UK commercial fast reactor design

    International Nuclear Information System (INIS)

    The design of the CDFR commercial demonstration fast reactor which should be put into operation early in the 90-ties is described. Basic elements of the reactor components are considered. The choice of the integrated primary coolant circuit, and design of intermediate heat exchangers, sodium pumps and charging machines is substantiated. The reactor power is 1320 MW(e), or 3300 MW(t). The sodium temperature at the reactor inlet is 370 deg C, at its outlet 540 deg C. Linear loading per fuel element length is 40 W/mm. The conclusion is drawn that the described design of the demonstration reactor fully corresonds to requirements of a full-scale commercial NPP with a fast reactor

  7. A glossary of terms for fast reactors

    International Nuclear Information System (INIS)

    The glossary aims to provide definitions of technical terms likely to be used in a fast reactor enquiry and to encourage the use of the same set of consistent terms in any documents intended for such an inquiry. In some cases definitions are formulated in the limited context of LMFBRS rather than applying to all types of reactors. A brief guide is presented to the different reactor types. (author)

  8. 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)

  9. Fast Reactor Fuel Development in Japan

    International Nuclear Information System (INIS)

    The future fast reactor and its fuel cycle system under development in Japan uses oxide fuel with simplified pelletizing fuel fabrication technology as a reference concept. Its driver fuel consists of large diameter annular fuel pellets, oxide dispersion strengthened ferritic steel cladding fuel pins with a ferritic-martensitic steel subassembly wrapper tube and minoractinide- bearing oxide fuel. The target burnup of the driver fuel is 150 GW.d/t in discharge average, which corresponds to 250 GW.d/t of peak burnup and 250 dpa of peak neutron dose. Fuel developmental efforts, including out-of-pile studies such as material characteristics experimental evaluation and fuel property measurements, various irradiation tests and fuel fabrication technology developments were planned and are in progress. Future fuels will be realized through Joyo irradiation tests and Monju demonstrations. International collaborative efforts are also an important part of such activities. (author)

  10. 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)

  11. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    that the cladding strength sufficient to withstand stress accounting for decreased thickness by the ACCI zone. (5) The wet wash and storage method was selected for disposing of the spent sodium bonded control rods, based upon experimental results at the JOYO facilities. The effects from storing sodium bonded control rods in wet storage were evaluated. The results indicated that these effect would not pose a safety problem. (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. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  14. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  15. 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)

  16. Commissioning of the cleaning loop of sodium contaminated equipment in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    The water vapor-nitrogen test loop was constructed to study the cleaning technology of sodium contaminated equipment in sodium-cooled fast reactor. The main components of the loop include a liquid nitrogen storage tank, an electric heating steam generator, a cleaning tank, an effluent storage tank, a condenser, a vacuum pump and a hydrogen meter. After a series of commissioning tests, some problems were found and the corresponding improvement measures were taken. The experiment procedure of the loop which was confirmed by the commissioning tests provided a guarantee for the following studies on the cleaning technology of sodium contaminated equipment. A number of experiences gained from the commissioning tests could be used in the equipment cleaning system of China Experimental Fast Reactor (CEFR). (authors)

  17. Status of sodium cooled fast reactor development in the Russian Federation

    International Nuclear Information System (INIS)

    This report describes the recent development and activities concerning fast reactors in Russia. The status of nuclear power in 1995 and operational experience of the BR-10, BOR-60 experimental reactors and of the BN-600 nuclear power plant are presented. Main results of R and D program in fast reactor area are discussed. (author)

  18. A Review of Fast Reactor Progress in Japan

    International Nuclear Information System (INIS)

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1980 through March 1981, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1979. The 1980 year budget for R&D work and for construction of a prototype fast breeder reactor, MONJU, will be approximately 14 and 19 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and an operational cycle at 75 MWt has been completed very recently. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor is being reviewed by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extraporating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized below

  19. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report

  20. Decommissioning of the Rapsodie fast reactor: developing a strategy

    International Nuclear Information System (INIS)

    The RAPSODIE experimental fast neutron reactor at Cadarache (France) was operated from 1962 to 1982. The initial decommissioning operations began immediately, reaching IAEA stage 2 in 1994. Since then, the facility has been maintained under surveillance pending final dismantling scheduled to begin in 2020. New studies are now in progress to accelerate the dismantling process. The present status of the reactor block is described, the advantages and drawbacks of early dismantling are considered, and various dismantling scenarios are discussed. (author)

  1. Thermomechanical analysis of fast-burst reactors

    International Nuclear Information System (INIS)

    Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor

  2. 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

  3. JNC viewpoint on fast reactor knowledge preservation

    International Nuclear Information System (INIS)

    JNC is undertaking a major program of research and development on liquid-metal cooled fast breeder reactors, which is fully supported by the government of Japan and the electrical utilities. Hence, the perspective of JNC on knowledge preservation is rather different from that of organizations where the fast reactor project has been scaled down or discontinued. Within JNC, there is a statutory obligation to preserve documentary records of the fast reactor project. Over time the method of archiving has changed from optical (microfilm, microfiche etc.) to digital storage. It is the long-term objective of JNC to convert all its records to digital format and make them available to staff over its intranet. JNC is also attempting to preserve 'human knowledge', that is, the expertise of staff who have been involved in the fast reactor project over a long period and who are now nearing retirement. Based on this information, two computerized systems are currently being constructed: one which records in,a readily accessible manner the background to key design decisions for the Monju plant; and a second which uses simple relationships between design parameters to aid designers understand the knock-on effects of design choices (joint project with Mitsubishi). To its partners in international cooperation - the US/DoE and the organizations of the Euro-Japan collaboration - JNC is proposing a joint approach to knowledge preservation and retrieval. The proposed concept, dubbed the International Super-Archive Network (ISAN), would make use of the standardized software the new technologies of the internet increase the mutual accessibility of fast reactor information. JNC considers it extremely important to reflect the lessons learnt from previous experience in the fast reactor field to the operation and maintenance of Monju and the design of future reactors. (author)

  4. Manufacture of Simulator of Irradiation Device for China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    This research belongs to the development of nuclear energy project of CEFR irradiation device design and previous studies of cladding material 316 (Ti) SS irradiation performance. The main content of the research is the development of irradiation

  5. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  6. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  7. Experimental reactor physics

    International Nuclear Information System (INIS)

    Neutronic experiments in moderators, subcritical assemblies, critical assemblies, and nuclear reactors are described, as well as the techniques of radiation measurements necessary to perform these experiments. Previously dispersed data from government reports, journal articles, and specialized monographs are codified. Original information drawn from the author's experience is included, especially on the pulsed source technique, spectrum measurements, research reactors, and exponential assemblies. The book provides the essential information for carrying out, analyzing, and understanding the experiments. Theory is kept to a minimum. Emphasis is placed on the physics of the situation, and the importance of estimating error as well as the mean value of a measured quantity

  8. Design challenges for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    It is of vital importance for commercialized fast reactor to achieve component design with excellent integrity and economics. In the phase II of feasibility study till 2005, a system design for commercialized fast reactor for sodium cooling was achieved. For economical improvement, the system design was undertaken along the guideline including innovative technology for system simplification and new material development. In this paper, the results from the design for shortening of cooling pipings, new components and three dimensional seismic isolation are described, which are design challenges for the sodium cooled fast reactor. Furthermore, in-service inspection and repair is mentioned. Finally, economics for the simplification and the mass reduction employing above technologies are examined

  9. 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)

  10. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    A general description of expert systems is given. The operation of a fast reactor is reviewed. The expert system to the diagnosis of breakdowns limited to the reactor core. The structure of the system is described: specification of the diagnostics; structure of the data bank and evaluation of the rules; specification of the prediagnostics and evaluation; explanation of the diagnostics; time evolution of the system; comparison with other expert systems. Applications to some cases of faults are finally presented

  11. Multi-group calculations for fast reactors

    International Nuclear Information System (INIS)

    The paper deals with various causes of error in calculations. The first part sets out the mathematical approximations (diffusion approximation, Sn method, etc.), the numerical resolution methods (effect of integration step), the models used, and the implications of these various factors in the determination of the principal characteristics of a fast neutron reactor. The second part studies the effect on reactivity of variations of element cross-sections, using various fuels, in a reactor of rather hard spectrum. (author)

  12. Removing the heat from fast reactor cores

    International Nuclear Information System (INIS)

    Whatever the view about the time when fast breeder reactors will reach the commercial and industrial stage, there is a growing and widespread interest in developing their technology. The reactors are called breeders because they can produce more fissile material than they use in their own cores. As part of an Agency programme related to their technology and economics a symposium on Alkali Metal Coolants - Corrosion Studies and System Operating Experience was held in Vienna from 28 November to 2 December

  13. 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...)

  14. Fast Reactor Knowledge Management at IGCAR, India

    International Nuclear Information System (INIS)

    The Process Architecture: → Acquire: Solicitation; Voluntary submission; Mandatory requirements; Interview/Observation; → Quality Control: Review/Editing; Certification; Quality index; → Disseminate: Publish through the Technology architecture; Formal/Informal Meetings; COPs; → Utilize: Projects; Day-to-day activities; → Maintenance; → Retirement. Mission: To conduct a broad based multidisciplinary programme of scientific research and advanced engineering development, directed towards the establishment of the technology of Sodium Cooled Fast Breeder Reactors (FBR) and associated fuel cycle facilities in the Country. The mission includes the development and applications of new and improved materials, techniques, equipment and systems for FBRs, pursue basic research to achieve breakthroughs in Fast Reactor technology

  15. Fast-acting nuclear reactor control device

    International Nuclear Information System (INIS)

    A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

  16. 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)

  17. International conference on fast reactors and related fuel cycles (FR09): Challenges and opportunities. Book of extended synopses

    International Nuclear Information System (INIS)

    Renewed interest in nuclear energy is driven by the need to develop carbon free energy sources, by demographics and development in emerging economies, as well as by security of supply concerns. It is expected that nuclear energy will deliver huge amounts of energy to both emerging and developed economies. However, acceptance of large scale contributions would depend on satisfaction of key drivers to enhance sustainability in terms of economics, safety, adequacy of natural resources, waste reduction, non-proliferation and public acceptance. Fast spectrum reactors with recycle enhance the sustainability indices significantly. This has led to the focus on fast spectrum reactors with recycle in the Generation IV International Forum (GIF) and the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative of the IAEA. It is expected that 2009 will register major events in the domain of fast spectrum reactors, that is, the restart of Monju in Japan, the first criticality of the China Experimental Fast Reactor in China, as well as new insights through end-of-life studies in Phenix, France. New fast reactors are expected to be commissioned in the near future: the 500 MW(e) Prototype Fast Breeder Reactor in India and the BN-800 unit in the Russian Federation. Moreover, China, France, India, Japan, Republic of Korea and the United States of America are preparing advanced prototypes/ demonstrations and/or commercial reactors for the 2020-2030 horizon. The necessary condition for successful fast reactor deployment in the near and mid-term is the understanding and assessment of innovative technological and design options, based on both past knowledge and experience, as well as on ongoing research and technology development efforts. In this respect, the need for in-depth international information exchange is underscored by the fact that the last large international fast reactor conference was held as far back as 1991. Since then, progress in research

  18. International conference on fast reactors and related fuel cycles (FR09): Challenges and opportunities. CN-176 presentations

    International Nuclear Information System (INIS)

    Renewed interest in nuclear energy is driven by the need to develop carbon free energy sources, by demographics and development in emerging economies, as well as by security of supply concerns. It is expected that nuclear energy will deliver huge amounts of energy to both emerging and developed economies. However, acceptance of large scale contributions would depend on satisfaction of key drivers to enhance sustainability in terms of economics, safety, adequacy of natural resources, waste reduction, non-proliferation and public acceptance. Fast spectrum reactors with recycle enhance the sustainability indices significantly. This has led to the focus on fast spectrum reactors with recycle in the Generation IV International Forum (GIF) and the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative of the IAEA. It is expected that 2009 will register major events in the domain of fast spectrum reactors, that is, the restart of Monju in Japan, the first criticality of the China Experimental Fast Reactor in China, as well as new insights through end-of-life studies in Phenix, France. New fast reactors are expected to be commissioned in the near future: the 500 MW(e) Prototype Fast Breeder Reactor in India and the BN-800 unit in the Russian Federation. Moreover, China, France, India, Japan, Republic of Korea and the United States of America are preparing advanced prototypes/ demonstrations and/or commercial reactors for the 2020-2030 horizon. The necessary condition for successful fast reactor deployment in the near and mid-term is the understanding and assessment of innovative technological and design options, based on both past knowledge and experience, as well as on ongoing research and technology development efforts. In this respect, the need for in-depth international information exchange is underscored by the fact that the last large international fast reactor conference was held as far back as 1991. Since then, progress in research

  19. TITAN program and direct cycle fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Yasuyoshi; Yoshizawa, Yoshio; Nitawaki, Takeshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

    2000-07-01

    In December 1999, the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (TIT) started a new program for the development of advanced nuclear reactors with small and medium size. TITAN is the acronym for the program. A novel concept of a carbon dioxide cooled direct cycle fast reactor with a Rankin cycle has been proposed as the advanced nuclear reactors and evaluated for an alternative option to liquid metal cooled fast reactors (LMFRs). The use of carbon dioxide as coolant eliminates major safety related problems of sodium cooled fast reactors: positive sodium void reactivity, hazardous reaction between sodium and water or air. The decay heat is passively removed by allocating a storage tank of liquidized carbon dioxide between the regenerator and the condenser, and by introducing naturally the carbon dioxide vaporized from the tank into the core in the event of the depressurization accident. The direct cycle results in considerable simplification of the heat transport system owing to the absence of intermediate cooling and water-steam loops comparing with the LMFRs. The thermal efficiency of the direct cycle is evaluated as 34.3 %, which is slightly higher than those in the current BWRs and PWRs. (author)

  20. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    A review of the United Kingdom Fast Reactor Programme is introduced. Operational experience with the Prototype Fast Reactor (PFR) is briefly summarized. The design concept of the Commercial Demonstration Fast Reactor (CDFR) is given in some detail. The emphasis is on materials development, chemical engineering/sodium technology, fuel reprocessing and fuel cycle, engineering component development and reactor safety

  1. 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)

  2. Plutonium utilization in thermal and fast reactors in Japan

    International Nuclear Information System (INIS)

    Nuclear power development in Japan is rather extensive, and the installed nuclear power capacity is predicted to reach 49,000MW(e) by 1985 and possibly 170,000MW(e) by 2000. Currently installed nuclear power is mainly based on the light-water reactor, and this trend is expected to persist for the time being. Plutonium produced by the LWR will reach 20t by 1985 and more than 200t by 2000. Should this plutonium be simply stored, it will cause economic pressure on utilities and the management, as well as physical protection problems associated with plutonium storing. Three ways of solving these problems are being worked out, the best solution being to use plutonium in fast reactors. To achieve this, an experimental fast reactor, JOYO, has been constructed and reached criticality in April 1977. A prototype fast breeder reactor, MONJU, designed for about 300MW(e), is nearing the final stages of design work. Its construction will commence in a few years. A demonstration fast breeder reactor will come after MONJU and the large-scale commercial use of a fast breeder reactor is expected around 1995. To meet the imminent need for plutonium utilization, two technologies, which are equally important to Japan, are currently being developed. One is the recycle use of plutonium into LWR, a technology which has long been jointly developed by research organizations and utilities. The other is to burn plutonium in an advanced thermal reactor (D2O-moderated, boiling-water cooled). The 160-MW(e) FUGEN is a prototype of this power reactor, and is almost finished. (author)

  3. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  4. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  5. Strengthening the R and D on fast reactor technology, and promoting its industrialization

    International Nuclear Information System (INIS)

    Based on the strategic thoughts of energy development in China expounded by Jiang Zemin in the article entitled 'Reflections on Energy Issues in China', the author points out in this paper that R and Ds on fast reactor technology shall be carried out timely in China by taking full advantage of international scientific resources, and overall planning in this regard shall be made as well. The point of view of strengthening fast reactor technology R and D and promoting its industrialization is also put forward in the paper. (authors)

  6. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    An examination has been made of how the currently available computing capabilities could be used to reduce Liquid Metal Fast Reactor design, manufacturing, and construction cost. While the examination focused on computer analyses some other promising means to reduce costs were also examined. (author)

  7. Use of fast reactors for actinide transmutation

    International Nuclear Information System (INIS)

    The management of radioactive waste is one of the key issues in today's discussions on nuclear energy, especially the long term disposal of high level radioactive wastes. The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) would allow 'burning' of the associated extremely long life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. The International Working Group on Fast Reactors (IWGFR) decided to include the topic of actinide transmutation in liquid metal fast breeder reactors in its programme. The IAEA organized the Specialists Meeting on Use of Fast Breeder Reactors for Actinide Transmutation in Obninsk, Russian Federation, from 22 to 24 September 1992. The specialists agree that future progress in solving transmutation problems could be achieved by improvements in: Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes); technological research and development on the design, fabrication and irradiation of the minor actinides (MAs) containing fuels; nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners; demonstration of the feasibility of the safe and economic MA burner cores; knowledge of the impact of maximum tolerable amount of rare earths in americium containing fuels. Refs, figs and tabs

  8. Fast reactor systems for deep sea research

    International Nuclear Information System (INIS)

    Fast reactor (FR) systems have been studied as power units for unmanned bases and research submersibles to monitor various phenomena and as a thermal source for the unmanned base to feed useful microorganisms in the deep sea region. The systems, which are set in pressure hulls, comprise of the FR's and secondary gas loops. Concepts and arrangements of the systems are presented. (author)

  9. Thermophysical properties of fast reactor fuel

    International Nuclear Information System (INIS)

    This paper identifies the fuel properties for which more data are needed for fast-reactor safety analysis. In addition, a brief review is given of current research on the vapor pressure over liquid UO2 and (U,PU)O/sub 2-x/, the solid-solid phase transition in actinide oxides, and the thermal conductivity of molten urania

  10. 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

  11. Development of plutonium: Fast Neutrons Reactors option

    International Nuclear Information System (INIS)

    Phenix reactor is shortly described with combustible assembly with some operational data. 'CAPRA'(Plutonium Enhance Consumption in Fast Reactors) is an R and D program for the development of an optimized combustible for fast reactors for burning more plutonium. Three ways are tested: a 45% Pu concentration in an oxide fuel keeping actual fabrication and reprocessing options giving a 80 kg/TWh Pu consumption, a fuel without U238 but with a W or a Mo matrix with problems of reprocessing and core reactivity giving a 110 kg/TWh Pu consumption, and a nitride fuel with an up to 65% Pu concentration giving a 90 to 100 kg/TWh Pu consumption. (A.B.)

  12. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Only minor changes were introduced in the French nuclear programme by the new government in 1981. The operating conditions of Rapsodie were very satisfactory up to January 1982. After a leak in the double primary jacket (nitrogen circuit) the reactor was shut down for investigations. Phenix is continuing to operate smoothly. Construction of Super Phenix (Creys Malville power plant) is proceeding normally though with some delay. The studies for the future (after Creys Malville) are following their way both for the Project 1500 (Super Phenix 2) and for the specific plants of the fuel cycle. Research and development are largely directed toward Super Phenix 1 needs and the prospects of Super Phenix 2. International cooperation remains very intensive. The financial resources devoted to the development of fast reactors are globally stable. Including fuel cycle and safety (but excluding the Phenix operation) about 1300 millions of francs will be devoted to fast reactors by the C.E.A. in 1982. (author)

  13. Fast Reactor Fuel Development in Europe

    International Nuclear Information System (INIS)

    Research and development of minor-actinide-bearing fuels in Europe has made significant progress, with a number of scoping irradiation tests made on a number of candidate fuels foreseen for fast reactors and dedicated minor actinide transmutation systems, e.g. the accelerator driven system. Currently, efforts concentrate on uranium based fuels, as the deployment of fast reactor fleets requires Pu generation in order to achieve sustainability. Both homogeneous and heterogeneous concepts for minor actinide reactor recycling are considered. In the former, the minor actinides are added in small quantities to the mixed oxide fuel, while in the latter, the minor actinides are loaded in significant quantities in UO2. Irradiation programmes to test these concepts for pellet and SPHEREPAC fuel configurations are under way. (author)

  14. Bowing and interaction of fast reactor subassemblies

    International Nuclear Information System (INIS)

    Deformations of the subassembly structural components, in particular the bowing of the hexagonal wrapper which encloses the pin bundles, due to stainless steel swelling as a result of fast neutron irradiation give rise to operational and safety problems especially in large breeder reactors where the neutron flux is much larger than in smaller reactors. The restraint on bowing induces heavy restraint loads and high stresses in the wrapper, which tend to limit the target burn-up of the fast reactor fuels. Therefore, a realistic analysis has to include the phenomenon of creep to determine the extent to which the stresses in the wrapper would be relaxed due to both thermally induced and irradiation induced creep. Apart from this, determination of deformations of the subassemblies in the core due to the interaction among them is also necessary. (author)

  15. Fast neutron spectrum determination with threshold detector at the RB reactor

    International Nuclear Information System (INIS)

    The fast neutron spectrum determination with threshold detectors at the RB reactor is described in this paper. The experimental results which are obtained on the coupled fast-thermal system CFTS-2 are given. At the end two different numerical methods for obtaining the fast neutron spectrum on the basis of experimental results are compared. (author)

  16. Axial distribution of absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    The coupled fast thermal system CFTS at the RB reactor is created for obtaining fast neutron fields. The axial distribution of fast neutron flux density in its second configuration (CFTS-2) is measured. The axial distribution of absorbed doses is computed on the basis of mentioned experimental results. At the end these experimental and computed results are given. (Author)

  17. Current status of work on fast reactors in the Union of Soviet Socialist Republics

    International Nuclear Information System (INIS)

    In the paper the status of work on fast reactors in the USSR as for the end of 1985 is presented. The problems of nuclear power development in the USSR, the design of BN-800 nuclear power plant, operating experience of the BN-600 nuclear power plant and of the BOR-60 experimental reactor, investigations at the BR-10 reactor, co-operation of socialist countries in the fast reactor area are considered

  18. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  19. Nuclear data needs for fast reactors

    International Nuclear Information System (INIS)

    The nuclear data, i.e., the numerical information about every nuclide - especially those representing the probabilities of various nuclear interactions and of radioactivity - of interest in a nuclear fission reactor are among the most essential inputs to be known a priori, to the best possible accuracy, for the design of nuclear reactor. The nuclides of interest cover not just (1) the fuel nuclides, the containers, the coolant, the moderator (if any), etc., that are initially inserted, but also (2) the actinides, the fission products, etc. that would be produced from the moment the reactor goes into operation and (3) the decay products that are produced even while the reactor is shutdown. The nuclide-list is known to cover a few hundreds. The neutron-nuclear interaction cross-section data, required for a few tens of reactions, very sensitively depend on the nuclide species and the neutron energy. Hence the data requirement significantly varies between thermal and fast reactors. The present talk is intended to touch upon the kinds and forms of nuclear data needed in the design and analysis of fast reactors. The recent variants available in the databases and some inter-comparison results will also be presented. (author)

  20. Fast Reactor Physics Parameters from a Pulsed Source

    International Nuclear Information System (INIS)

    One of the more important integral parameters in fast reactor physics analysis is the neutron spectrum of a particular composition reactor core. Various methods, such as proton recoil counters and nuclear emulsion analysis, have been used to study fast reactor spectra. With the development of high intensity short-duration pulsed neutron sources, the time-of-flight technique has become suitable for fast reactor spectrum determination. To evaluate the feasibility of measuring fast neutron spectra from a core using time-of-flight techniques, an experiment has been performed to measure the equilibrium spectmm in a large block of depleted uranium using the General Atomics Linac facilities. A ten-metric-ton block of depleted uranium was assembled to form a 81-cm cube. This block of uranium was pulsed by electron bombardment of a uranium target imbedded in the block. The spectra from various sections of the block were measured using time-of-flight techniques for a 50-m flight path. Spectral indices, such as the ratio of the fission rates of U238/U235, U233/U235, U234/U235, Np237/U235, Pu239/U235 were also measured. In addition, measurements of the U238 capture rates were obtained in various parts of the block. This paper describes the techniques used to obtain these reactor physics parameters. The experimental results such as the spectra and spectral indices are also compared with those obtained from theoretical considerations using multigroup transport theory analysis. The pulsed neutron technique is also applicable for the measurement of such parameters as: β/ℓ, where β is the effective delayed neutron fraction and ℓ is the lifetime; neutron importance; and keff. This paper concludes with a discussion on the proposed application of a pulsed neutron source for the measurement of some of these parameters on fast reactor cores constructed on ZPR-VI, the Argonne Fast Critical Facility. (author)

  1. Super fast reactor R and D projects in Japan

    International Nuclear Information System (INIS)

    The Japanese research project of the 'Research and Development of Super Fast Reactor' was conducted from December 2005 to March 2010, entrusted by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT). Aiming at a highly economical fast reactor, the plant concept was developed with quantitative characteristics. The databases of the thermal hydraulics and materials including water chemistry were accumulated by experiments. Based on the success of the project, the second phase of Super FR project was initiated in July 2010. The project consists of three subjects; (1) development of the plant concept: (2) thermal-hydraulics: (3) material-coolant interactions: The Super Fast Reactor has the same plant system of the once-through coolant cycle as the Super Light Water Reactor, the thermal reactor. The results of experimental R and D constitute the common database for the development of Super LWR and Super FR. This paper describes the principle of the reactor concept development and the R and D program of the second phase project. (author)

  2. Overview of Lead Based Reactor Design and R&D Status in China

    International Nuclear Information System (INIS)

    Liquid lead or lead based alloy is a potential candidate coolant for fast reactors and accelerator driven system (ADS) subcritical reactors because of its many unique nuclear, thermophysical and chemical attributes. The Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead based reactors. A series of China lead based Reactors (named CLEAR) design, the lead based experimental loops (KYLIN series Pb–Bi loops and DRAGON series PbLi loops), a high intensified D-T neutron generator (HINEG) and structure material (CLAM) were developed by the Institute of Nuclear Energy Safety Technology. In this paper, the CLEAR design and R&D activities are presented. (author)

  3. Synthesis of the first experimental results obtained on the PROFIL-R and M experiments performed in the Phenix fast neutron reactor

    International Nuclear Information System (INIS)

    The PROFIL-R (fast spectrum) and PROFIL-M (moderated spectrum) experiments were performed between 2003 and 2008 in the French fast neutron reactor Phénix. These experiments consisted of the irradiation of pure isotope samples in a well-characterized neutrons flux in order to collect accurate information on the total capture integral cross sections of the principal heavy isotopes and some important fission products in the spectral range of fast reactor. This method can be used for all isotopes transformed by neutron capture into a stable or long-lived nuclide and is based on the measurement of the composition change induced by irradiation. Therefore, accurate and reproducible measurements of isotopic compositions and concentrations of the elements (actinides and fission products) before and after irradiation are required. The major difficulty for the analyses of these actinides and fission products is the low quantity of the initial powder enclosed in steel container (3 to 5 mg) and the very low quantities of products formed (several μg) after irradiation. We present the developments performed during the last few years by laboratories of the French Commission on Atomic Energy and Alternative Energies (CEA) in order to acquire very accurate and precise isotopic and elemental data on selected irradiated powders. Among the necessary developments are the conceptions of systems set in shielded hot-cells to open the steel containers and collect the full amount of powders, and the set-up of specific analytical methods for mass spectrometry measurements in order to obtain isotopic and elemental ratios at uncertainty of few per mil level. A synthesis of the results obtained and first preliminary interpretations will also be presented. (author)

  4. Advanced sodium fast reactor unit concept

    International Nuclear Information System (INIS)

    The paper presents status of development for 1200 MW power unit with sodium fast reactor for commercial construction in the Russian Federation. General characteristics of the reactor plant (RP) and power unit as well as goals that shall be achieved because of design development are described. The power unit design is based on technical decisions, which have been partially proven during sodium reactor operation in Russia and partially have been validated by R and D work for BN-800 RP. At the same time, new technical decisions are applied that improve safety and technical-and-economic indices. To validate them, the corresponding R and D work shall be performed. It is planned to construct the pilot power unit in 2020 and to put into operation the next commercial power units of this type using plutonium generated in the thermal reactors. (author)

  5. 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

  6. Knowledge management in fast reactors and related fuel cycles

    International Nuclear Information System (INIS)

    Full text: The 21st century is ushering in a new phase of economic and social development which can be referred as 'Knowledge Economy', in which knowledge has become the key asset in determining the organization's success or failure. The IAEA defines knowledge management as an integrated, systematic approach to identify, manage and share an organization's knowledge collectively in order to help achieve the objectives of the organization. Nuclear technology is very complex and a highly technical endeavor. It relies on innovative creation, storage and dissemination of knowledge. The nuclear energy is characterized by long time scales and technological excellence. Nuclear knowledge management is a critical input to nuclear power industry, the associated fuel cycle activities and nuclear applications in medicine, industry and agriculture. Realizing the importance of knowledge preservation in the area of fast reactor technology, IAEA had given a consultancy work to Argonne National Laboratory to study and suggest the means of knowledge management. The IAEA initiative seeks to establish a comprehensive inventory of fast reactor data and knowledge for the fast reactor development in the coming years. It was suggested that the knowledge regarding important disciplines like fuels and materials, reactor physics and core design, operations, the demonstration of safety should be preserved. Various countries have initiated the fast reactor knowledge preservation activities. In France, CEA, EDF and Framatome ANP have initiated liquid metal cooled fast reactor knowledge preservation project that deals with R and D aspects and Superphenix design. European Fast Reactor collaboration (MASURCA,SNEAK,ZEBRA) has preserved the zero power critical experimental data in the SNEDAX database. Japan has started a comprehensive knowledge preservation program including the capture of 'Human Knowledge' based on interviews. In Russia steps are initiated to preserve fast reactor knowledge

  7. Review of the Italian fast reactor programme

    International Nuclear Information System (INIS)

    The Caorso power station (860 MWe) underwent its third reloading at the end of 1985. The construction work on the power station at Montalto di Castro, a twin 1,000 MWe BWR reactor plant has continued according to plan. ENEL has agreed with Ansaldo the contract for the NSSS of the Trino Vercellese nuclear plant. This will be the first of the so-called ''Unified Design Nuclear Power Station'', twin PWR units, Westinghouse type, 950 MWe each, with 3 cooling loops. The value of the NSSS order is 1400 billion lire, while the complete cost will be 5000 billion lire. Civil engineering work will begin in July 1987 and completion is planned for 1995. In the fast reactor field, the Italian effort has been operating in the framework of the European R and D agreement; during 1985 the sum assigned by ENEA to fast reactors, excluding PEC realization, was about 100 billion lire. Worthy of note was the signing of a formal agreement between ENEA and ENEL with the aim of coordinating their activities on fast reactor development

  8. Fast Pyrolysis of Agricultural Wastes in a Fluidized Bed Reactor

    Science.gov (United States)

    Wang, X. H.; Chen, H. P.; Yang, H. P.; Dai, X. M.; Zhang, S. H.

    Solid biomass can be converted into liquid fuel through fast pyrolysis, which is convenient to be stored and transported with potential to be used as a fossil oil substitute. In China, agricultural wastes are the main biomass materials, whose pyrolysis process has not been researched adequately compared to forestry wastes. As the representative agricultural wastes in China, peanut shell and maize stalk were involved in this paper and pine wood sawdust was considered for comparing the different pyrolysis behaviors of agricultural wastes and forestry wastes. Fast pyrolysis experiments were carried out in a bench-scale fluidized-bed reactor. The bio-oil yieldsof peanut shell and maize stalk were obviously lower than that ofpine sawdust. Compared with pine sawdust, the char yields of peanut shell and maize stalk were higher but the heating value of uncondensable gaswas lower. This means that the bio-oil cost will be higher for agricultural wastes if taking the conventional pyrolysis technique. And the characteristic and component analysis resultsof bio-oil revealed that the quality of bio-oil from agricultural wastes, especially maize stalk, was worse than that from pine wood. Therefore, it is important to take some methods to improve the quality of bio-oilfrom agricultural wastes, which should promote the exploitation of Chinese biomass resources through fast pyrolysis in afluidized bed reactor.

  9. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  10. The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

    International Nuclear Information System (INIS)

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)

  11. 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)

  12. 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

  13. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    CERN Document Server

    Lee, Seung Kyu; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spectra with high radiation from a gamma-ray background. The signals from neutrons and gamma-rays are discriminated by the digital charge pulse shape discrimination (PSD) method, which uses total to partial charge ratio analysis. The signals are digitized by a flash analog-to-digital convertor (FADC). To evaluate the performance of the fabricated stilbene neutron diagnostic system, the efficiency of 10 mm soft-iron magnetic shielding and the detection efficiency of fast neutrons were tested experimentally using a 252Cf neutr...

  14. Fast Food McDonald's China Fix

    Institute of Scientific and Technical Information of China (English)

    DAVID HENDRICKSON

    2006-01-01

    @@ Since the opening of its first outlet 16 years ago, McDonald's China operation has on many levels proven enormously successful.Home to more than 750 locations nationwide, the Middle Kingdom today ranks as one of McDonald's ten largest markets,with returns hovering in doubles digits and raking in billions annually. As lucrative as it may be, however, China has nonetheless developed into a relative sore spot for the world's leading fast food giant.

  15. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  16. 3 Investment Scenarios for Fast Reactors

    International Nuclear Information System (INIS)

    Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

  17. Safety problems in fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Fast neutron reactors fuels have a high proportion of plutonium and undergo severe irradiation. Risks during spent fuel reprocessing and subsequent fabrication will depend on isotopic composition, fission product content, physico-chemical form of products, quantities handled. These risks (criticality, contamination, irradiation) are listed for the different steps of the cycle and methods used to control the risks (chemical reaction yields, equipment reliability, intervention, conditions...) are indicated. Problem arising from wastes and effluents produced at each step are briefly given

  18. The current status of utilization of research reactors in China

    International Nuclear Information System (INIS)

    Seminars on utilization of research reactors were held to enhance experience exchanging among institutes and universities in China. The status of CARR (China Advanced Research Reactor) project is briefly described. The progress in BNCT program in China is introduced. (author)

  19. Design study of high breeding fast reactor

    International Nuclear Information System (INIS)

    Aiming to increase fuel breeding capability as the most essential feature of fast breeders, an idea of the FP gas purge/tube-in-shell type metallic fuel assembly is proposed. It makes volume fraction of fuel high as more than 50% and realizes a very hard neutron spectrum in the core. The structure of the fuel assembly, its fabrication and the FP gas purging mechanism were assessed and it is clarified that the new concept of the fuel assembly is engineeringly feasible. FP gas purging does not affect shielding structure and can be managed by a small scale cover-gas treatment system because of good trapping characteristics of bonding sodium in the assembly as expected. The fuel handling system without forced cooling is possible. Other reactor components such as IHX were also evaluated. Thus, a concept of the total reactor system of a fast breeding reactor of 670 MWe with the ultra-high breeding ratio of 1.84 and the short reactor doubling time of 6.7 years was obtained. (author)

  20. Opening Address [FR09: International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, Kyoto (Japan), 7-11 December 2009

    International Nuclear Information System (INIS)

    Full text: Distinguished guests, ladies and gentlemen. It is my honour to address participants at this opening session of the International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, organized by the IAEA and hosted by the Japan Atomic Energy Agency. Fast reactor technology has the potential to ensure that energy resources which would last hundreds of years with the technology we are using today will actually last several thousand years. In other words, it can satisfy enormous increases in demand. This innovative technology also reduces the risk to the environment and helps to limit the burden that will be placed on future generations in the form of waste products. The coming year will be an exciting one for the development of fast spectrum nuclear reactors. We expect to reach several important milestones: (a) The first criticality of the China Experimental Fast Reactor; (b) The restart of the Monju prototype fast reactor in Japan; (c) The new insights we will gain through the end-of-life studies at the Phenix reactor in France. In the near future, new fast reactors will be commissioned: the 500 MW(e) Prototype Fast Breeder Reactor in India, the first in a series of five of the same type, and the BN-800 reactor in the Russian Federation. Moreover, China, France, India, Japan and the Republic of Korea are preparing advanced prototypes and demonstration or commercial reactors for the 2020-2030 period. Nuclear power is set to become an increasingly important part of the global energy mix in the coming decades as demand for energy grows. A number of countries in both the developed and developing world have told the IAEA that they are interested in introducing nuclear power. The 30 countries which already have nuclear power reactors are set to build more. This trend is likely to be accompanied by accelerated deployment of fast reactors. Continued advances in research and technology development are necessary to ensure improved

  1. A review of the fast reactor programme in Japan

    International Nuclear Information System (INIS)

    In Japan the experimental reactor ''Joyo'' has provided abundant experimental data and excellent operational records attaining 40,000 hours operation in total by the end of 1989, since its first criticality in 1977. On the prototype reactor ''Monju'', more than eighty percent of construction work has already been completed on schedule, aiming at the initial criticality by October 1992. As for the demonstration fast breeder reactors (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee. (author). Figs and tabs

  2. Recent progress of Gas Fast Reactor program

    International Nuclear Information System (INIS)

    The GFR is considered by the French Atomic Energy Commission as a promising concept which combines the benefits of fast spectrum and high temperature, using helium as coolant. He properties are interesting with respect to safety: it is single phase (no threshold effect due to phase changing), chemically inert, and non toxic. It affords an optical transparency allowing potential improvements in temperature measurement, management for dismantling, and in-service-inspection. The voiding effect is limited, less than 1$, providing quasi- decoupling of the reactor physics from the state of the coolant. Nevertheless, Helium is a poor coolant, so that the GFR viability includes development of a refractory and dense fuel, and robust management of accidental transients, especially cooling accidents. GFR feasibility is essentially linked to three demonstrations: the feasibility (fabrication, thermo-mechanical behaviour) of a refractory fuel; the safety architecture with appropriate systems for the prevention and a robust mitigation of accidental scenarios (especially depressurization); economic competitiveness. The first one includes an experimental activity at the laboratory scale: completion of the results is expected by 2012-2015. The next step afterward will be the design, construction and the operation of a 50-100 MWth experimental reactor, the Allegro project (former ETDR), possibly as a European Joint Undertaking. The full paper will recall the 2007 design choices and it will give an overview of the progress performed so far regarding the safety architecture and the safety evaluation. The 2007 reference fuel technology is a ceramic plate type fuel element. It combines a high enough core power density (minimization of the Pu inventory), plutonium and minor actinides recycling capabilities. Innovative to many aspects, the fuel element is a key issue in the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is

  3. European lead fast reactor-ELSY

    International Nuclear Information System (INIS)

    Highlights: → ELSY, the European Lead Fast Reactor (LFR) design is presented. → Presentation of Main Components design. → Core design, safety systems and safety analysis. → Future development activities for Lead-cooled system. - Abstract: The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the EU-FP6-ELSY project. The ELSY (European Lead-cooled System) reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, while fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative Decay Heat Removal (DHR) systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators

  4. Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs

  5. Risk Management for Sodium Fast Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

  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. Validity of using UPuO2 vibropack experimental fuel pins in reactors on fast and thermal neutrons: First experiments on conversion of weapons grade plutonium into nuclear fuel

    International Nuclear Information System (INIS)

    Extensive scope of scientific and technological work has been carried out in SSC RF RIAR to substantiate usage of vibropack oxide fuel pins in fast and thermal neutron reactors. In fulfilling the work, physical-mechanical and technological characteristics of granulated fuel have been studied, radiation tests and material science investigations of mock-up, experimental and research fuel pins of BN-type (in BOR-60 and BN-600 reactors) and WWER-1000 type (in SM-2 and MIR reactors) have been carried out. Total quantity of fabricated fuel pins is about 30 000 pieces. In BOR-60 reactor, maximum burn-up attained 30% h.a. for regular SA and burnup was of 32,3% h.a. for experimental fuel pins of the dismantled SA. In testing UPuO2 vibropack fuel pins in BN-600 reactor, maximum burn-up of -10,8% h.a. was attained. Post irradiation examinations of fuel pins have revealed that since the problems of both chemical and thermo-mechanical fuel-cladding interactions have been solved, the resource of the fuel pins like these would only depend on the choice of cladding material. Vibropack fuel pins, containing UPuO2 under conditions of MIR reactor attained burn-up more than 30 MW day/kg U both under nominal operation and under load-following modes. The experience in designing, manufacturing and operating the facilities on fabrication of granulated uranium and MOX fuel and fuel pins is gained. The data bank and calculation codes, describing vibropack fuel pin behavior under different operation modes is created. According to the Concept of RF Minatom on recovery of surplus weapon-grade plutonium, resulting from disarmament, the State Scientific Center of Russian Federation RIAR (Dimitrovgrad) has begun a practical realization of the technology on conversion of metal weapon-grade plutonium into mixed uranium-plutonium oxide fuel. Processing has been carried out and granulated UPuO2 fuel for BOR-60, BN-600 reactors and experimental batches of granulated fuel for mock-up and experimental

  8. Substantiation of physical concepts of fast reactors in Russia: experience and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

    2008-07-01

    The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

  9. 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.)

  10. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    This report describes the development and activities on fast reactor in Japan for the period of April 1996 - March 1997. During this period, the 30th duty cycle operation has been started in the Experimental Fast Reactor ''''Joyo''''. The cause investigation on the sodium leak incident has completed and the safety examination are being performed in the Prototype Fast Breeder Reactor ''''Monju''''. The three years design study since FY1994 on the plant optimization of the Demonstration FBR has been completed by the Japan Atomic Power Company (JAPC). Related research and development works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of Power Reactor and Nuclear Fuel Development Corporation (PNC), JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). In November 1996, the Japan Atomic Energy Commission (JAEC) established a Social Gathering Meeting to discuss generally the significance of FBR development in Japan for the future. (author)

  11. Advanced Fast Reactor - 100 - Design Overview

    International Nuclear Information System (INIS)

    The Advanced Fast Reactor-100 (AFR-100) is a small modular sodium-cooled fast reactor with an electrical power output of 100MWe. The AFR-100 has a long-lived core that does not require refueling for 30 years. The concept contains various innovations such as a small compact modular core (both vented and non-vented fuel pins), advanced core shielding materials, a compact fuel handling system, advanced electromagnetic pumps, compact intermediate heat exchangers, and a direct reactor auxiliary cooling system. These advanced systems and components were adopted in order to reduce the overall size of the primary heat transport system (and therefore the overall commodities and cost), enhance safety, and to improve overall plant performance. This paper presents the summary results of a year-long study that culminated in the design of two primary heat transport configurations for the AFR-100. The paper describes those innovations and shows how they are integrated into the overall AFR-100 primary heat transport system design. (author)

  12. Sodium fast reactors (SFRs) and recyclers

    International Nuclear Information System (INIS)

    This presentation is about Sodium Fast Reactor (SFRs) and Recyclers. Their pursuit has been going on in the United States (U.S.) since 1941 and that development work could help support the penetration of SFRs into the current nuclear power market in three forms: 1. A breeding SFR to increase the supply of fissile material. It will not happen for many decades because of increased uranium (U) resources, nuclear market ability to absorb increased U prices, and/or switch to a Thorium (Th) fuel cycle (under development in India) until the anticipated stringent regulations for breeding SFRs are defined and tested. 2. An economic SFR capable of competing with the Advance Light Water Reactor (ALWR) expected to produce electricity in the near future. The Generation IV (Gen IV) program is pursuing that goal under conceptual studies in South Korea (1) and, particularly under the demonstration Japan Sodium Fast Reactor (JSFR) (2) forecasted to start up by 2025 followed by the deployment of commercial JSFRs before 2050. 3. To use the pyro-processing and electro refining methodology developed under the Integral Fast Reactor (IFR) (3) to separate the Light Water Reactor (LWR) spent nuclear fuel (SNF) Transuranics (TRUs) and to burn them in SFRs referred to as Advanced Burner Reactors (ABR). That innovative approach can significantly increase the capacity of geological repositories for disposition of LWR SNF. That last form of SFR is needed urgently to cope with the continued increase in U.S. inventories of recyclable fissile and fertile materials and, particularly, with the projected growth in LWR SNF. According to a recent Electrical Power Research Institute (EPRI) study (4) to reduce CO2 emissions, the U.S. nuclear generated electricity will increase by 64 Gigawatt electrical (GWe) by 2030. While it is realized that additional long term interim storage can alleviate this need, it is not a long term solution because it will have to be followed eventually by final disposal or

  13. Designs characteristics, and development of fast reactors for utilization of thorium

    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 a smooth development of liquid metal fast breeder reactors in Brazil is proposed and designs and characteristics of required reactors are discussed. Emphasis is placed on utilization of thorium that is abundant in the country, on reactor safety in order to promote public acceptance and smoothness of the development. The initial step is the construction of a 5 MW experimental reactor in order to acquire basic experiences and technologies. The second step is the construction of a series of small power reactors designed with particular emphasis on safety and ease of operation. In the final phase when fast breeder reactors are to play a central role in electricity generation, large power reactors that utilize both uranium and thorium fuel cycles will be built to establish a practically permanent power system. (Author)

  14. A review of fast reactor program in Japan - April 1984

    International Nuclear Information System (INIS)

    The fast breeder reactor development project in PNC has been in progress steadily in these eighteen years. Concerning the experimental fast reactor, JOYO, the MK-II core attained criticality on November 22, 1982 with 51 fuel assemblies, and received the ''Certificate of Inspection before Operation'' from Government Authority on March 31, 1983, after 100 hours operation with the rated output of 100 MW. Since then, the core has been utilized to implement irradiation bed characteristics test, and to irradiate fuels and structural materials especially for the prototype reactor MONJU. With respect to the prototype reactor MONJU, the installation permit was issued on May 27, 1983, from the prime minister, and the contracts of the first stage between PNC and fabricators were made recently. At the same time, almost all the licenses of preparatory construction works were issued by March 1983, and preparatory construction works were started in April 1983. On the other hand, conceptual design of a demonstration reactor is now under way in a close cooperation with concerned authorities and utilities, as well as investigations of the way of conducting necessary research and development

  15. Sodium fast reactor power monitoring using gamma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.; Normand, S.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A.M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, CEA - Saclay DRT/LIST/DETECS/SSTM, Batiment 516 - P.C. no 72, Gif sur Yvette, F-91191 (France); Montagu, T.; Dautremer, T.; Barat, E. [CEA, LIST, Laboratoire Processus Stochastiques et Spectres (France); Ban, G. [ENSICAEN (France)

    2009-06-15

    This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the

  16. Sodium fast reactor power monitoring using gamma spectrometry

    International Nuclear Information System (INIS)

    This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the signal. An

  17. Principle results of design and experimental-test works on space nuclear energy facility on thermoemission reactor-transformer base on fast neutrons

    International Nuclear Information System (INIS)

    At the S. P. Korolev Rocket-Space Corporation (RSC) 'Energy' for long time pilot and research works on space Nuclear Energy Facilities (NEF) of electric capacity from 150 to 500 kW and MW range in nuclear electric-rocket engine facilities (NEREF) of its base have been carried out. Peculiarity of Nuclear Energy Facility is application of thermal emission reactor-converter (TRC) on fast neutrons, application of pure lithium-7 isotope in the capacity of coolant, and the high-temperature niobium alloy in the capacity of construction materials, module design of TRC and NEF at all. Principal attention is paid to basing of nuclear and radiation safety for production, launching, commissioning and decommissioning of these facilities. At present at the RSC 'Energy' the main volume of pilot work is emphasized on the module scheme NEF with TRC of electric capacity from 150 kW to 800-1000 kW for electric power plants creation for both inhabited Lunar base, and mine and processing complex for industrial Moon development. In the paper different possibilities for effective use of NEF and NEREF uniform technologies for further space development are given

  18. Fast critical experiment data for space reactors

    International Nuclear Information System (INIS)

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, the old experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, we have calculated about 20 fast benchmark critical experiments with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors

  19. Fuel systems for compact fast space reactors

    International Nuclear Information System (INIS)

    About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom density, operating temperatures, fuel swelling, fission gas release, fuel-cladding compatibility, and consequences of failure. It was concluded that UO2 and UN fuels show approximately equal performance potential and that UC fuel has lesser potential. W/Re alloys have performed quite well as cladding materials, and Ta, Nb, and Mo/Re alloys, in conjunction with W diffusion barriers, show good promise. Significant issues to be addressed in the future include high burnup swelling of UN, effects of UO2-Li coolant reaction in the event of fuel pin failure, and development of an irradiation performance data base with prototypically configured fuel pins irradiated in a fast neutron flux

  20. Fast reactor technology innovation and visualization

    International Nuclear Information System (INIS)

    Innovations in safety, operations, and maintenance for improving the availability, reliability, and capital cost of the sodium fast reactor are described. Concerning safety these innovations deal with on-line limiting safety settings, inherent core protection, detection of subassembly coolant mis-allocation. Concerning reactor operations these innovations deal with advanced energy conversion, adapting non-base load nuclear plants and on-line diagnostics. Other innovations concern inspection, servicing, refueling. The development of these innovations rely on visualization technology for their use and for demonstration of improvements achievable. A visualization platform for running these innovations and the nuclear plant thermal-hydraulic, structure, and process codes that underlie them are described. The platform hardware consists of a large-scale tiled display and a haptic hand-controller and in the future will grow to include a high-speed network and multiple graphics-client systems

  1. Chemistry for fast reactor fuel cycle

    International Nuclear Information System (INIS)

    The fuel cycle for the fast reactors poses several challenging chemistry issues. The use of fuels with high plutonium content, the variety of fuel matrices (oxides, carbides, metal alloys), the high burn-up to which the fuel is driven and the need to close the fuel cycle with minimum out-of-pile inventory are examples of special features of fast reactors. The need to reduce waste generation and the need to identify matrices for safe long term disposal of waste are additional issues that need a chemist's attention. As a chemist, the subject of actinide separations has been very stimulating to me, with a myriad of interesting possibilities and at the same time, demanding careful attention to the unique chemistry of the actinides including multiplicity of oxidation states. The presence of high concentrations of plutonium in the reprocessing streams introduces issues such as third phase formation, which provides an incentive for the development of candidates for solvent extraction as alternatives to tri-n-butyl phosphate, currently used for the Purex reprocessing scheme. With the advent of supercritical fluid extraction as a tool for actinide recovery from a variety of matrices, and the potential of room temperature ionic liquids to offer significant advantages in actinide processing, actinide separations is an element of fast reactor fuel cycle that is full of opportunities and challenges. The need to process metallic alloy fuels using molten salt electrorefining as the route, adds further to the challenges. The presentation will highlight some of the recent progress achieved in this area at IGCAR. (author)

  2. Fast Reactor Programme. Third Quarter 1969. Progress Report

    International Nuclear Information System (INIS)

    The RCN research programme on fast spectrum nuclear reactors comprises reactor physics, fuel performance, radiation damage in canning materials, corrosion behaviour in canning materials, aerosol research and heat transfer and hydraulics. An overview is given of the fast reactor experiments at the STEK critical facility in Petten, the Netherlands, in the third quarter of 1969

  3. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

  4. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    An interdisciplinary systemic approach to socio-technical optimization of nuclear energy management is proposed, by recognizing a) the rising requirements to nuclear safety being realized using fast reactors (FR), b) the actuality to maintain and educate qualified workforce for fast reactors, c) the reactor safety and public awareness as the keystones for improving attitude to implement novel reactors. Knowledge management and informational support firstly is needed in: 1) technical issues: a) nuclear energy safety and reliability, b) to develop safe and economic technologies; 2) societal issues: a) general nuclear awareness, b) personnel education and training, c) reliable staff renascence, public education, stakeholder involvement, e).risk management. The key methodology - the principles being capable to manage knowledge and information issues: 1) a self-organization concept, 2) the principle of the requisite variety. As a primary source of growth of internal variety is considered information and knowledge. Following questions are analyzed indicating the ways of further development: a) threats in peaceful use of nuclear energy, b) basic features of nuclear risks, including terrorism, c) human resource development: basic tasks and instruments, d) safety improvements in technologies, e) advanced research and nuclear awareness improvement There is shown: public education, social learning and the use of mass media are efficient mechanisms forming a knowledge-creating community thereby reasoning to facilitate solution of key socio-technical nuclear issues: a) public acceptance of novel nuclear objects, b) promotion of adequate risk perception, and c) elevation of nuclear safety level and adequate risk management resulting in energetic and ecological sustainability. (author)

  5. Actinide management with commercial fast reactors

    Science.gov (United States)

    Ohki, Shigeo

    2015-12-01

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GWey if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  6. The Argentine-Brazilian fast reactor programme

    International Nuclear Information System (INIS)

    This paper summarizes the Argentine-Brazilian Fast Reactor Programme and gives reasons for the decision of a binational venture. The work carried out by both countries is described, showing how they complement each other, with the corresponding saving of resources. The main objectives of the Programme and tentative schedules in three progressing integrating stages are given and the present nuclear know-how in each country is identified as a good starting point. The paper also gives some details regarding the economical and human resources involved. (author). 1 graph

  7. Actinide management with commercial fast reactors

    International Nuclear Information System (INIS)

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GWey if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel

  8. Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves

  9. Actinide management with commercial fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ohki, Shigeo [Japan Atomic Energy Agency, 4002, Narita-cho, O-arai-machi, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2015-12-31

    The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GW{sub e}y if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

  10. Small power sodium cooled fast nuclear reactors

    International Nuclear Information System (INIS)

    1.5 MW(e), 12 MW(e) and 170 MW(e) small power sodium cooled fast reactors have been developed. The reactor plants were developed as universal power units for economically effective energy and industrial steam generation and heat supply. The main features increasing the power unit economic efficiency are: serial fabrication of standard RPs at the factory and delivery of reactor vessels in ready made form; realization of self-protection principles and use of passive systems in RP; use of standard machine room equipment, fabricated in accordance with the rules of conventional heat power engineering; use of turbine plant with thermodynamic coefficient, exceeding the corresponding value for the plants of PWR type. For MBRU-1.5 and MBRU-12 RPs it is proposed to use a core without FA replacement during the whole service life (30 years) and for BMN-170 RP it is proposed to use a core with a 4 year operating period and 1 year between the refueling shutdowns. During the whole service life a minimal number of operating personnel will be needed for the plant servicing. The personnel functions will be periodically to observe the parameters of technological process. Passive principles are used in the main RP safety systems: a passive type system of emergency residual heat removal system provides heat removal directly through the reactor vessel forced air cooling due to the natural air chimney effect; an emergency reactor shut-down system is provided by emergency protection rods with active-passive action. (author)

  11. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  12. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  13. History of fast reactor fuel development

    Science.gov (United States)

    Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

    1993-09-01

    The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

  14. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    International Nuclear Information System (INIS)

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  15. Fast reactors in Russia: Status as of 2000 and prospects

    International Nuclear Information System (INIS)

    intellectual) resources. Therefore, great attention is paid in Russia to coordination and integration of international efforts in the development of nuclear technologies. At the UN Millennium Summit on September 6, 2000, the President of the Russian Federation announced the initiative on organization within the framework of an international project with IAEA participation and development of innovative reactor technology and nuclear fuel cycle with natural safety eliminating proliferation of nuclear weapons and providing incineration of plutonium and other long lived radioactive elements. Establishment of a special IAEA group on the innovative nuclear reactors and fuel cycles aimed at the analysis, choosing and development of advanced nuclear technology, is the first step in this direction. Fast neutron reactor technology is most promising from the standpoint of meeting imposed requirements. Undoubtedly, new advanced nuclear technologies will be chosen on the basis of results achieved in the existing technologies. Therefore, it is important to involve experts from other IAEA working groups, including TWGFR, in discussions and examination of advanced options. Three fast reactors are in operation in Russia in 2000: Test reactor BR-10, experimental reactor BOR-60 and prototype reactor BN-600. Current status and basic areas of design studies of fast reactor technology are described. Research related to accelerator driven systems in Russia include experimental studies of accelerator driven system parameters and analytic studies, and computer codes development

  16. A small modular fast reactor as starting point for industrial deployment of fast reactors

    International Nuclear Information System (INIS)

    The current commercial reactors based on light water technology provide 17% of the electricity worldwide owing to their reliability, safety and competitive economics. In the near term, next generation reactors are expected to be evolutionary type, taking benefits of extensive LWR experience feedbacks and further improved economics and safety provisions. For the long term, however, sustainable energy production will be required due to continuous increase of the human activities, environmental concerns such as greenhouse effect and the need of alternatives to fossil fuels as long term energy resources. Therefore, future generation commercial reactors should meet some criteria of sustainability that the current generation cannot fully satisfy. In addition to the current objectives of economics and safety, waste management, resource extension and public acceptance become other major objectives among the sustainability criteria. From this perspective, two questions can be raised: what reactor type can meet the sustainability criteria, and how to proceed to an effective deployment in harmony with the high reliability and availability of the current nuclear reactor fleet. There seems to be an international consensus that the fast spectrum reactor, notably the sodium-cooled system is most promising to meet all of the long term sustainability criteria. As for the latter, we propose a small modular fast reactor project could become a base to prepare the industrial infrastructure. The paper has the following contents: - Introduction; - SMFR project; - Core design; - Supercritical CO2 Brayton cycle; - Near-term reference plant; - Advanced designs; - Conclusions. To summarize, the sodium-cooled fast reactor is currently recognized as the technology of choice for the long term nuclear energy expansion, but some research and development are required to optimize and validate advanced design solutions. A small modular fast reactor can satisfy some existing near-term market niche

  17. Seismic response analysis of the PEC fast reactor building

    International Nuclear Information System (INIS)

    In order to compute the motion induced by the design earthquakes at the vessel supporting structure, a seismic response analysis was performed for the PEC fast reactor, taking into account the effects of soil-structure interaction by use of experimentally determined soil parameters. The main aim of he analysis was to evaluate the safety margins present in the design calculations. A detailed 3D finite element model was set up for fixed base analysis; from the results of the 3D model a simplified equivalent model of the structure was then derived for soil-structure interaction analysis. The mathematical model was validated and calibrated by using the results of the experimental dynamic tests performed on the reactor building. The results have shown the adequacy of the computation methodologies, and in particular of those on the use of the equivalent model. (author)

  18. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  19. Delayed gamma power measurement for sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Graphical abstract: Display Omitted Research highlights: →20F and 23Ne tagging agents are produced by fast neutron flux. →20F signal has been measured at the SFR Phenix prototype. → A random error of only 3% for an integration time of 2 s could be achieved. →20F and 23Ne power measurement has a reduced temperature influence. → Burn-up impact could be limited by simultaneous 20F and 23Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,α) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  20. 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)

  1. Technical meeting to 'Preserve fast reactor physics knowledge'. Working material

    International Nuclear Information System (INIS)

    The meeting extended its scope beyond reactor physics to include all the main areas of fast reactor data retrieval and knowledge preservation (FR KP). The participants presented the status of the national FR KP efforts and the progress achieved since the kick-off meeting of the IAEA initiative (meeting hosted by ANL-West in Idaho Falls, Idaho, 2-4 April 2002). Details are given in Section 2. The Scientific Secretary of the Technical Working Group on Fast Reactors (TWG-FR) presented the Agency activities (KNK II data and documentation retrieval and preservation), and recalled the Agency's role in this initiative: - Coordination of the national efforts - Ensuring the collaboration with other International Organizations (mainly OECD/NEA) - Establishing and maintaining the access means to the ultimate goal of the initiative, the 'fast reactor knowledge base'. The integration of specific activities relevant to the FR KP initiative, which are planned within the framework of the TWG-FR, was discussed. It was agreed to implement the following as TWG-FR tasks with clear relevance to FR KP initiative: - Japanese 'Proposal from Monju relevant to Fast Reactor Knowledge Preservation Activity in the framework of the IAEA TWG-FR' - Proposal of a CRP on 'Generalization and Analyses of Operational Experience with Fast Reactor Equipment and Systems' - TM on 'Handling of Sodium Coming from Decommissioned Fast Reactors and from the Shutdown of Experimental Facilities' (if not already covered by the TECDOC being prepared by IAEA's Nuclear Waste Technology Section). While the responsibility for fast reactor knowledge preservation, data retrieval and interpretation, as well as quality assurance will rest with the individual Member States joining the FR KP initiative, the participants confirmed the Agency's role (see above). More specifically, the participants in the meeting recommended that the IAEA - support and coordinate data retrieval and interpretation efforts by the fast reactor

  2. 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

  3. Seminar on Heat-transfer fluids for fast neutron reactors

    International Nuclear Information System (INIS)

    This book reports the content of a two-day meeting held by the Academy of Sciences on the use of heat-transfer fluids in fast neutron reactors. After a first part which proposes an overview of scientific and technical problems related to these heat-transfer fluids (heat transfer process, nuclear properties, chemistry, materials, risks), a contribution proposes a return on experience on the use of heat-transfer fluids in the different design options of reactors of fourth generation: from mercury to NaK in the first fast neutron reactor projects, specific assets and constraints of sodium used as heat-transfer fluid, concepts of fast neutron reactors cooled by something else than sodium, perspectives for projects and research in fast neutron reactors. The next contribution discusses the specifications of future fast-neutron reactors: expectations for fourth-generation reactors, expectations in terms of performance and of safety, specific challenges. The last contribution addresses actions to be undertaken in the field of research and development: actions regarding all reactor types or specific types as sodium-cooled reactors, lead cooled reactors, molten salt reactors, and gas-cooled fast reactors

  4. LFR "Lead-Cooled Fast Reactor"

    Energy Technology Data Exchange (ETDEWEB)

    Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C; Suh, K

    2006-05-11

    The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small, transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology

  5. Status of Phenix operation and of sodium fast reactors in the world

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Martin, L. [Phenix plant, 30 - Bagnols sur Ceze (France); Courtois, C. [CEA Marcoule 30 (France)

    2007-07-01

    The French fast breeder reactor (FBR) Phenix restarted in 2003 after 6 years of safety reevaluation procedures. The goal of the experiments performed at Phenix is, first, to demonstrate the technical feasibility of transmutation of minor actinides and long-life products in a fast reactor and secondly, to acquire knowledge on structure materials for future energy systems and on innovative nuclear fuel concepts. After several years of Generation IV discussions, many countries have announced or confirmed their priority for the fast sodium reactor as a reference design. These countries today include Japan, China, Korea, India and Russia (simultaneously with lead reactors). The United States have announced a project for a waste-burning reactor. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision of building a prototype scheduled for operation in 2020. These declarations are all sustained in a very practical manner by ongoing events in this field. Following the excellent results obtained by the BN-600 (600 MWe), Russia has re-launched the BN-800 project. China is currently in the process of building a 75 MWt research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU (250 MWe) for divergence in 2008. In India, a 1200 MWt power reactor is under construction, scheduled for divergence in 2010, the first of 3 planned sodium reactors.

  6. Indian fast reactor technology: Current status and future programme

    Indian Academy of Sciences (India)

    S C Chetal; P Chellapandi

    2013-10-01

    The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

  7. The economics of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Specific technical features of fast reactors make them more expensive than water-cooled reactors in terms of initial investment, an over cost of 30% is acknowledged in this study. Their consumption of natural uranium is negligible being fed on depleted uranium (except for the very first cycle when an important quantity of plutonium is necessary). In the context of the scarcity of natural uranium, fast reactors could provide a competitive KWh compared with PWR. The study shows that sodium-cooled fast reactor could be economically competitive somewhere in the second part of the 21. century. The development of fast reactors could be accelerated by other arguments than economic competitiveness, for instance some governments might value more the energy independence given by a fleet of fast reactors or by considerations linked to non-proliferation or to the burning of actinides. In addition the article details the worldwide resource in natural uranium. (A.C.)

  8. Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

  9. Sodium fast neutron reactors. Status and perspective of development

    International Nuclear Information System (INIS)

    This report reveals data on development history of domestic fast neutron reactors cooled with sodium (BN reactors). It also shows BN reactors' unique role in expanding source of nuclear power raw materials and in solving ecological problems relating to radioactive wastes. There is brief information on characteristics and operation experience of research reactors BR-10, BOR-60, pilot-industrial reactors BN-350 and BN-600. As well there is data on BN-800 reactor designing that obtained a license for building. There are considered BN reactor peculiarities in regard of safety and design decisions on safety provision at the level meeting standard document requirements. BN reactor technical and economic indices and the ways of their improvement are evaluated. There is brief information on alternative perspective technologies of fast reactors, in particular regarding 'BREST-300' reactor cooled with lead coolant

  10. Trial visualization of fast reactor design knowledge

    International Nuclear Information System (INIS)

    In design problems of large-scale systems like fast breeder reactors, inter-relations among design specifications are very important where a selected specification option is transferred to other specification selections as a premise to be taken account in engineering judgments. These inter-relations are also important in design case studies with the hypothetical adoption of rejected design options for the evaluation of deviation propagations among design specifications. Some of these rejected options have potential worth for future reconsideration by some circumstance changes (e.g., advanced simulations to exclude needs for mock-up tests, etc.), to contribute to flexibility in system designs. In this study, a computer software is built to visualize a design problem structure by representing engineering knowledge nodes on individual specification selections along with inter-relations of design specifications, to validate the knowledge representation method and to derive open problems. (author)

  11. 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)

  12. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  13. Technological problems in the use of research fast reactors for radiotherapy of patients with malignant tumors

    International Nuclear Information System (INIS)

    The authors discuss the technological problems associated with the use of fast neutrons in radiotherapy of cancer patients and outline the approaches to the solution of these problems. The state of the art is assessed. Physical and radiobiologial prerequisites for the use of fast reactors for radiotherapy of patients with malignant tumors are analyzed. Results of clinic used of BR-10 reactor at the Medical Radiology Research Center, Russian Academy of Medical Sciences, are presented. Experimental and clinical findings indicate that the results of radiotherapy may be appreaciably improved if a novel perspective source of fast neutrons, a nuclear reactor, is used

  14. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  15. Behavior of actinides in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides (237Np, 240Pu, 241Am, and 243Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors' confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  16. Actinide behavior in the Integral Fast Reactor. Final project report

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides (237Np, 240Pu, 241Am, and 243Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  17. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mixed oxide (MOX) (U,Pu)O2, and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO2), where PuO2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO2/PuO2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  18. Methane reforming with fast nuclear reactor steam

    International Nuclear Information System (INIS)

    The paper considers the concept of utilizing nuclear fast reactor (FR) with a sodium coolant for methane steam reforming. Steam conditions of a power FR, e.g. the BN-600 now operating in Russia: steam pressure P=13.2 MPa and steam temperature T=500degC, do not absolutely comply with the catalytic reactor working parameters, which produces a synthetic gas (syngas), a mix of hydrogen and carbon oxide. In this connection, the present paper addresses a possibility of utilizing steam produced in one of three independent the BN-600 loops in an amount of 640 t/h for preparing a gas-steam mixture with T=500degC and its additional heating in a converter up to the operating temperature, T=850degC, at the expense of natural gas burning or electrical energy supplying. In this case, the fraction of burned natural gas burning or electrical energy supplying. In this case, the fraction of burned natural gas significantly decreases. It is estimated that steam parameters of the BN-600 afford to obtain ∼3·105 nm3/h of hydrogen. It is also considered a concept of nuclear heat transfer to remote regions to be achieved with the aid of syngas incoming from the converter, its cooling further and transmitting through a pipeline to the place of its utilization, where it is restored into methane with the heat extraction. (author)

  19. Nuclear data requirements for fast and intermediate reactor calculations

    International Nuclear Information System (INIS)

    During 1960/61, some work has been done at Karlsruhe in the compilation of reliable nuclear data for fast- and intermediate-reactor calculations. Materials included thus far are He, O16, Na23, Cr, Fe, Ni, Mo, U235, U238 and Pu239. For fast- and intermediate-reactor optimization studies, reliable theoretical prediction of space-dependent neutron energy spectra, critical masses, breeding ratios, burn-up, Doppler coefficients and related subjects, the author was interested in much more detailed microscopic cross-section data concerning mainly neutron absorption and fission than are needed, for example, for the evaluation of few-group constants in the conventional multi-group diffusion theory. In spite of much progress in the experimental determination and theoretical interpretation of nuclear data, many inconsistencies in the experimental results of different authors and laboratories and large gaps in experimental nuclear-data work still remain. This paper discusses a great number of these gaps and inconsistencies in respect of the above-mentioned nuclei. (author)

  20. Fast neutron reactions and fast neutron flux in the NRX reactor

    International Nuclear Information System (INIS)

    This report deals with fast neutron reactions and fast neutron flux in NRX. By fast neutrons it is meant those neutrons having an energy from ∼ 1 Mev to ∼ 25 Mev. The report is divided into three parts. In the first part measurements of (n,2n) cross sections in different irradiation positions of NRX are described. It is shown that from these experimental data, any (n,2n) cross section in NRX can be estimated. In the second part, measurements of the fast neutron flux in different irradiation positions of NRX are described. In the third part the values of the fast neutron flux found in Part II are used to estimate a variety of (n,p) and (n,a) cross sections in NRX. Fast neutron reactions in a reactor have been studied by many different workers. However, this report is not intended to give an exhaustive bibliography on fast neutron research; it will only refer the reader to a few publications where a great deal of information on fast neutron reactions as well as references to earlier work can be found, There is an excellent chapter on fast neutron research in the book of Hughes on Pile Neutron Research (Hughes 1953). A tabulation of the published measurements of threshold reactions for fission neutrons has appeared in Nucleonics (Rochlin 1959), Mellish et al. (1958) have discussed at length flux and cross section measurements with fast neutrons. The reader is also referred to recent work on this subject published in the Canadian Journal of Physics (Roy at el. 1958; Eastwood and Roy 1959; Roy 1959). (author)

  1. Design and layout decision for refueling system of advanced fast neutron reactors

    International Nuclear Information System (INIS)

    Describes fast neutron reactor refueling features, BN-1200 power unit general data, its refueling system design concepts, individual refueling equipment purpose and designs, and required experimental studies to create it. Refueling equipment characteristics for BN-800 and BN-1200 reactors are compared. (author)

  2. Measurement of fast neutron spectra inside reactors with a Li6 semiconductor counter spectrometer

    International Nuclear Information System (INIS)

    The possibility of using the Li6 semiconductor counter spectrometer for measuring fast neutron spectra inside reactors has been investigated in details and some solutions of the difficulties associated with the high interference of thermal neutrons in well-moderated reactors are suggested and checked experimentally (author)

  3. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  4. History of fast reactor fuel development

    International Nuclear Information System (INIS)

    The first fast breeder eactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s. (orig.)

  5. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  6. A review of fast reactor progress in Japan

    International Nuclear Information System (INIS)

    The fast reactor development project in Japan is continuing at a slightly increased scale of effort in budget. The total budget for LMFBR development for fiscal year 1978 was 24 billion yen. In August 1977 major industries engaged in LMFBR have set up an office where design work can be jointly conducted. Highlights and topics of the fast reactor development activities cover description of JOYO reactor, its first criticality experiment, and the prototype fast breeder MONJU. Research and development programmes dealt with fission products release and its possible interaction with the soodium coolant, inspection of reactor components, experiments simulating sodium leakage, development of steam generator

  7. Materials development and materials selection for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    As applied to operational conditions of fast reactors a combined investigations of structural materials are accomplished. Steels 10Kh18N9, 08Kh16N11M3 are recommended to be used for reactor vessels, steel 10Kh2M - for steam generators, a strain hardened steel 08Kh16N11M3T and steel 05Kh12N2M - for fuel assembly cans. The investigations provided for designing such sodium cooled fast reactors as Bor-60, BN-350, BN-600. The investigation results are now in use in construction of a new fast reactor BN-800

  8. Fast reactor core management in Japan: twenty years of evolution at JOYO

    International Nuclear Information System (INIS)

    Twenty years of operations at the experimental fast reactor JOYO provide a wealth of experience with core and fuel management. This experience has been applied to several core modifications to upgrade JOYO's irradiation capability. Core physics tests and Post Irradiation Examination (PIE) results have been used to confirm the accuracy of neutron diffusion theory calculations. These experiences and accumulated data will be useful for the core design in future fast reactors in Japan's development. (author)

  9. Absorbed neutron doses in air holes of fast neutron fields at the RB reactor

    International Nuclear Information System (INIS)

    Different experimental fast neutron fields are created at the RB reactor. The absorbed neutron doses in their air holes are determined on the basis of intermediate and fast neutron spectra measurements. The obtained results are analyzed in connection with application of these fields. (author)

  10. Multigroup fast fission factor treatment in a thermal reactor lattice

    International Nuclear Information System (INIS)

    A multigroup procedure for the studies of the fast fission effects in the thermal reactor lattice and the calculation of the fast fission factor was developed. The Monte Carlo method and the multigroup procedure were combined to calculate the fast neutron interaction and backscattering effects in a reactor lattice. A set of probabilities calculated by the Monte Carlo method gives a multigroup spectrum of neutrons coming from the moderator and entering the fuel element. Thus, the assumptions adopted so far in defining and calculating the fast fission factor has been avoided, and a new definition including the backscattering and interaction effects in a reactor lattice have been given. (author)

  11. A CFD Simulation Process for Fast Reactor Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Kurt D. Hamman; Ray A. Berry

    2010-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k–e and SST (Menter) k–? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  12. A CFD simulation process for fast reactor fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hamman, Kurt D., E-mail: Kurt.Hamman@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Berry, Ray A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States)

    2010-09-15

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly 'benchmark' geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-{epsilon} and SST (Menter) k-{omega} were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  13. Fast reactors bulk sodium coolant disposal NOAH process application

    International Nuclear Information System (INIS)

    Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

  14. Implications of Fast Reactor Transuranic Conversion Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

    2010-11-01

    Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (“burners”) do not have blankets; the cases above CR=1 (“breeders”) have breeding blankets. The burnup was allowed to float while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is “attractive” for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found

  15. Fast reactor nuclear physics parameters calculation code system 'EXPARAM'

    International Nuclear Information System (INIS)

    The calculation code system ''EXPARAM'' was designed to analyze the experimental results systematically measured at the fast critical assembly (FCA) in Tokai research establishment of JAERI. Some calculation codes developed independently in JAERI and in US research institutes were collected and arranged as the fast reactor physics calculation code system. The multi-group core calculation code and the perturbation calculation code based on the diffusion theory and transport theory calculate the physics parameters such as eigenvalue, reaction rate, Doppler reactivity worth and sodium void worth. The dynamic physics parameters such as prompt neutron lifetime and effective delayed neutron fraction are also calculated. Input and Output data of calculation codes are transferred to each other using a direct access file on UNIX computer system. (author)

  16. Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

    International Nuclear Information System (INIS)

    After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

  17. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.

  18. Development of studies on helium cooled fast reactors

    International Nuclear Information System (INIS)

    A necessity is shown of developing breeders with high reproductive properties. Helium cooled fast reactor is considered. The reactor performances, heating circuit with the use of a steam turbine unit in the secondary circuit is outlined. The reactor design and fuel assemblies are described

  19. 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

  20. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  1. Model of fast reactor knowledge preservation system

    International Nuclear Information System (INIS)

    Despite lack of the energy market today, fast reactors (FR) in the closed nuclear fuel cycle are the basis of a full-scale development of nuclear power in future. However, there are serious problems concerning the future R and D of these reactor technologies related to the following obstacles. All research on FR was stopped in Germany, Italy, United Kingdom and the United States and the work performed only dealt with the decommissioning of FR. Many experts who participated in R and D programs to create FR have retired or are approaching retirement age. In France, Japan and Russia work on the development of FR still continues, but there is a lack of young scientists and engineers. Due to all this factors IAEA launched the initiative to combine efforts of the leading nuclear countries to develop a project for the preservation of knowledge in the field of scientific and technological problems of FR development. Efforts of IAEA and national experts resulted in a model of FR information search and classification (so called ). This work has initiated a systematic process of creation and filling of information data bank on various aspects of FR design and operation. As the next step it would be logical to develop self-consistent mathematical models of FR-based NPP and closed NFC with their subsequent introduction into the system of knowledge preservation. So, it will serve as an important step towards preservation of knowledge in the field of FR design through joint development and to ensure open access to software. Such a project may lay the groundwork for the future development of distance learning courses and training on the optimal FR design, with the participation of leading specialists in this field. The report provides a mathematical and logical model for the preservation of knowledge concerning FR science and technology: taxonomy, an engineering model of FR-based NPP, a FR NFC model

  2. The search for advanced remote technology in fast reactor reprocessing

    International Nuclear Information System (INIS)

    Research and development in fast reactor reprocessing has been under way ∼ 20 yr in several countries. During the past decade, France and the United Kingdom have developed active programs in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the Experimental Breeder Reactor II (EBR-II) facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. The Federal Republic of Germany (FRG) and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper focuses on the search for improved facility concepts and better maintenance systems in the CFRP, and, in turn, on how developments at ORNL have influenced the technology elsewhere

  3. IAEA fast reactor knowledge preservation initiative. Project focus: KNK-II reactor, Karlsruhe, Germany

    International Nuclear Information System (INIS)

    This Working Material (including the attached CD-ROM) documents progress made in the IAEA's initiative to preserve knowledge in the fast reactor domain. The brochure describes briefly the context of the initiative and gives an introduction to the contents of the CD-ROM. In 2003/2004 a first focus of activity was concentrated on the preservation of knowledge related to the KNK-II experimental fast reactor in Karlsruhe, Germany. The urgency of this project was given by the impending physical destruction of the installation, including the office buildings. Important KNK-II documentation was brought to safety and preserved just in time. The CD-ROM contains the full texts of 264 technical and scientific documents describing research, development and operating experience gained with the KNK-II installation over a period of time from 1965 to 2002, extending through initial investigations, 17 years of rich operating experience, and final shutdown and decommissioning. The index to the documents on the CD-ROM is printed at the end of this booklet in chronological order and is accessible on the CD by subject index and chronological index. The CD-ROM contains in its root directory also the document 'frclassification.pdf' which describes the classification system used for the present collection of documents on the fast reactor KNK-II

  4. Immobilization of Fast Reactor First Cycle Raffinate

    International Nuclear Information System (INIS)

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method

  5. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

  6. Immobilization of Fast Reactor First Cycle Raffinate

    Energy Technology Data Exchange (ETDEWEB)

    Langley, K. F.; Partridge, B. A.; Wise, M.

    2003-02-26

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

  7. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  8. A review of the U.K. fast reactor programme: March 1978

    International Nuclear Information System (INIS)

    The review of the UK fast reactor programme covers the description of Dounreay Fast Reactor shut down after seventeen years of successful operation; description of prototype fast reactor (PFR); core design parameters safety features and plant design for commercial demonstration fast reactor (CDFR). Engineering development is related to large sodium rigs, coolant circuit hydraulics and vibration, instrumentation and components. The subjects of interest are material development, sodium technology, fast reactor fuel, fuel cycle, reactor safety, reactor performance studies

  9. Technology Options for a Fast Spectrum Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

    2006-06-01

    Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

  10. 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

  11. 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.)

  12. Preparation of covariance data for the fast reactor. 2

    International Nuclear Information System (INIS)

    For some isotopes important for core analysis of the fast reactor, covariance data of neutron nuclear data in the evaluated nuclear data library (JENDL-3.2) were presumed to file. Objected isotopes were 10-B, 11-B, 55-Mn, 240-Pu and 241-Pu. Physical amounts presumed on covariance were cross section, isolated and unisolated resonance parameters and first order Legendre coefficient of elastic scattering angle distribution. Presumption of the covariance was conducted in accordance with the data estimation method of JENDL-3.2 as possible. In other ward, when the estimated value was based on the experimental one, error of the experimental value was calculated, and when based on the calculated value, error of the calculated one was obtained. Their estimated results were prepared with ENDF-6 format. (G.K.)

  13. Preparation of covariance data for the fast reactor. 2

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Keiichi; Hasagawa, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    For some isotopes important for core analysis of the fast reactor, covariance data of neutron nuclear data in the evaluated nuclear data library (JENDL-3.2) were presumed to file. Objected isotopes were 10-B, 11-B, 55-Mn, 240-Pu and 241-Pu. Physical amounts presumed on covariance were cross section, isolated and unisolated resonance parameters and first order Legendre coefficient of elastic scattering angle distribution. Presumption of the covariance was conducted in accordance with the data estimation method of JENDL-3.2 as possible. In other ward, when the estimated value was based on the experimental one, error of the experimental value was calculated, and when based on the calculated value, error of the calculated one was obtained. Their estimated results were prepared with ENDF-6 format. (G.K.)

  14. Calculation capability of NETFLOW++ code for natural circulation in sodium cooled fast reactor

    International Nuclear Information System (INIS)

    The present paper describes the simulation of the natural circulation in the secondary heat transport system (HTS) after an intentional plant trip of the experimental fast reactor 'Joyo' with the 140 MWt irradiation core using the plant dynamics analysis code NETFLOW++. This code is an integrated network code to calculate the nuclear steam supply system (NSSS) and the balance of the plant (BOP), i.e., turbine/feedwater system. Up to now, the code has been validated using transient data of the experimental sodium facility PLANDTL, experimental fast reactor 'Joyo' and the prototype fast breeder reactor 'Monju'. These validations are steps to evaluate the natural circulation transient of a large-scale fast breeder reactor. Therefore, the former validation results are introduced to show the degree of agreement. In order to consolidate the applicability of the code to the evaluation of the natural circulation, the present test was selected and simulated using the NETFLOW++ code. Major plant parameters are simulated with good agreement such a similar accuracy as the Mimir-N2 exclusive code for 'Joyo'. As a result, it is concluded that the NETFLOW++ is applicable to the natural circulation analysis of sodium-cooled fast reactors with the similar scale of the prototype reactor 'Monju'. (author)

  15. Fast reactor development programme in France during 1992

    International Nuclear Information System (INIS)

    The present position with respect to the development of fast reactors in France and prospects for future R and D is summarized. The paper gives an overview on the status of the fast reactors Phenix and Super Phenix. In addition to the studies in support of the EFR project, which are presented in a separate report, CEA and NOVATOME have conducted exploratory studies to evaluate the potential of fast reactors to burn plutonium and long lived wastes with the objective to maintain the acceptable values of two important parameters for safety, namely the sodium void worth and the Doppler coefficient. (author). 1 fig

  16. Accuracy of helium accumulation fluence monitor for fast reactor dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Chikara; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    A helium (He) accumulation fluence monitor (HAFM) has been developed for fast reactor dosimetry. In order to evaluate the measurement accuracy of neutron fluence by the HAFM method, the HAFMs of enriched boron (B) and beryllium (Be) were irradiated in the Fast Neutron Source Reactor `YAYOI`. The number of He atoms produced in the HAFMs were measured and compared with the calculated values. As a result of this study, it was confirmed that the neutron fluence could be measured within 5 % by the HAFM method, and that met the required accuracy for fast reactor dosimetry. (author)

  17. SAMOFAR - a paradigm shift in reactor safety with the molten salt fast reactor

    International Nuclear Information System (INIS)

    SAMOFAR - Safety Assessment of the Molten Salt Fast Reactor - is a 5M€ project of the European Union research program Horizon 2020. The project consortium consists of 11 participants and the fundamental research part is mainly executed by universities and research laboratories, like CNRS, JRC, ClRTEN, TU Delft and PSI, thereby exploiting each other's unique expertise and infrastructure. The grand objective of SAMOFAR is to prove the innovative safety concepts of the Molten Salt Fast Reactor (MSFR) by advanced experimental and numerical techniques, to deliver a breakthrough in nuclear safety and optimal waste management, and to create a consortium of stakeholders to demonstrate the MSFR beyond SAMOFAR. Furthermore, we will build a software simulator to demonstrate the operational transients, and we will show the mild responses of the MSFR to transients and accident scenarios, using new leading-edge multi-physics simulation tools including uncertainty quantification. All experimental and numerical results will be incorporated into the new reactor design, which will be subjected to a new integral safety assessment method

  18. Designs and Experiments for Studies of Fast Neutron Fields at the RB Reactor

    International Nuclear Information System (INIS)

    The RB reactor is a heavy water critical assembly that has been in operation since 1958 using, at different times, natural metal uranium, 2% enriched metal uranium, and 80% enriched aluminium dioxide fuel of Soviet origin. A feasibility study of the RB reactor as a fast neutron source began in 1976, and four versions of fast neutron fields around or in the reactor were designed through 1990: an external neutron converter (ENC) in 1976; an experimental fuel channel (EPC) in 1982, an internal neutron converter (lNC) in 1983, and a coupled fast-thermal core (HERBE) in 1990. This paper presents an overview of the characteristics and experimental applications of each particular fast neutron field mentioned above, including available irradiation space, neutron spectra, and equivalent neutron and gamma dose rates. Control and safety-related implications of these modifications are emphasized. The computer codes and nuclear data libraries used in calculations are described briefly. (author)

  19. Emergency cooling down of fast-neutron reactors by natural convection (a review)

    Science.gov (United States)

    Zhukov, A. V.; Sorokin, A. P.; Kuzina, Yu. A.

    2013-05-01

    Various methods for emergency cooling down of fast-neutron reactors by natural convection are discussed. The effectiveness of using natural convection for these purposes is demonstrated. The operating principles of different passive decay heat removal systems intended for cooling down a reactor are explained. Experimental investigations carried out in Russia for substantiating the removal of heat in cooling down fast-neutron reactors are described. These investigations include experimental works on studying thermal hydraulics in small-scale simulation facilities containing the characteristic components of a reactor (reactor core elements, above-core structure, immersed and intermediate heat exchangers, pumps, etc.). It is pointed out that a system that uses leaks of coolant between fuel assemblies holds promise for fast-neutron reactor cooldown purposes. Foreign investigations on this problem area are considered with making special emphasis on the RAMONA and NEPTUN water models. A conclusion is drawn about the possibility of using natural convection as the main method for passively removing heat in cooling down fast-neutron reactors, which is confirmed experimentally both in Russia and abroad.

  20. A new safety approach in the design of fast reactors

    International Nuclear Information System (INIS)

    A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

  1. Fast reactor fuel pin behaviour modelling in the UK

    International Nuclear Information System (INIS)

    Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

  2. Analysis of fast reactor scenario with different conversion ratios

    International Nuclear Information System (INIS)

    Korean fast reactor scenarios have been analyzed for various kinds of conversion ratio by the DANESS system dynamic analysis code. The once-through fuel cycle analysis was modeled based on the Korean 'National Energy Basic Plan' up to 2030 and a postulated nuclear demand growth rate until 2150. The fast reactor scenario analysis has been performed for three kinds of conversion ratios such as 0.3, 0.61 and 1.0. Through the calculations, the nuclear reactor deployment scenario, front-end cycle, back-end cycle, and long-term heat load have been investigated. From the once-through results, it is shown that the nuclear power demand would be ∼70 GWe and the total amount of the spent fuel accumulated by 2150 would be ∼168000 t. Also, the fast reactor scenario analysis results show that the spent fuel inventory and out-pile transuranic element can be reduced by increasing the fast reactor conversion ratio. Furthermore, the long-term heat load of spent fuel decreases with increasing the conversion ratio. However, it is known that the deployment of a fast reactor of low conversion ratio does not much reduce the spent fuel and out-pile transuranic element inventory due to the fast reactor deployment limitation which is related to the availability of transuranic elements. (author)

  3. Coupled hydro-neutronic calculations for fast burst reactor accidents

    International Nuclear Information System (INIS)

    Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

  4. 20F power measurement for generation IV sodium fast reactors

    International Nuclear Information System (INIS)

    The Phenix nuclear power plant has been a French Sodium Fast Reactor (SFR) prototype producing electrical power between 1973 and 2010. The power was monitored using ex-core neutron measurements. This kind of measurement instantly estimates the power but needs to be often calibrated with the heat balance thermodynamic measurement. Large safety and security margins have then been set not to derive above the nominal operating point. It is important for future SFR to reduce this margin and working closer to the nominal operating point. This work deals with the use of delayed gamma to measure the power. The main activation product contained in the primary sodium coolant is the 24Na which is not convenient for neutron flux measurement due to its long decay period. The experimental study done at the Phenix reactor shows that the use of 20F as power tagging agent gives a fast and accurate power measurement closed to the thermal balance measurement thanks to its high energy photon emission (1.634 MeV) and its short decay period (11 s). (authors)

  5. Dynamic behavior of the Fast Reactor cores: the Symphony program

    International Nuclear Information System (INIS)

    A fast reactor core is schematically constituted of Fuel Assemblies and Neutronic Shields, immersed in the primary coolant (sodium) which circulates inside the assemblies. Two main physical phenomena have a strong influence on the dynamic behavior of this system: the impacts between the beams and the interactions with the fluid. The impacts between the beams limit the relative displacements. The fluid leads to “inertial effects”, with globally lower vibration frequencies, and “dissipative effects”, with higher damping. Symphony is an important research program on the seismic behaviour of the fast reactor cores, developed from 1993 to 1998 at the CEA Saclay, with both experimental and theoretical parts. The experiments are at a representative scale, with Fuel Assemblies (or FA) and Neutronic Shields (or NS). Test are made “in air” (without fluid) and “in water”, to study the influence of the fluid (the sodium). A numerical model has been built for the interpretation of the tests. The interpretation of the tests is made by using a simple and efficient numerical method, based on the Euler equations for the fluid and homogenization techniques, which yields low computational costs. Impacts between the beams are taken into account also. The gaps between the feet and the grid plate lead to high damping for the beams if the gaps are important. The fluid leads to a strong coupling between the FA and the NS in the whole core, and limits the relative displacement. (author)

  6. Experimental spectra unfolding of fast ion backscattering

    International Nuclear Information System (INIS)

    Problems on processing of experimental spectra of fast light ion backscattering are considered to obtain information about element composition in thin films and surface layers of solids. Application of mathematical processing of the spectra is shown to allow considerably to improve analytical characteristics of the ion backscattering method and to expand the field for its application

  7. Mitigation of gas entrainment in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Argon cover gas may entrain into sodium in the hot pool and in the surge tank of liquid Sodium cooled Fast Reactor (SFR) due to various mechanisms. The entrained cover gas may hinder the normal reactor operation in many ways such as reduction in heat transfer in the heat exchanger, causing neutronic perturbation inside the core etc. The recirculating gas bubbles also enhance chances of cavitation in the pumps. Therefore, it is required to mitigate gas entrainment in reactor. High free surface velocity and turbulence level are the causes of gas entrainment from free surface. Hence, gas entrainment can be avoided by reducing these factors at free surface. This is achieved by employing gas entrainment mitigation devices which essentially alters the flow pattern and reduces velocity and turbulence level at the free surface. A combined experimental and computational approach is proposed to develop such devices. The computational model which is used for parametric studies is first validated against experiments carried out in scale down water model of reactor primary circuit and surge tank. The effect of different gas entrainment mitigating devices for reduction in free surface velocity and turbulence level has been analyzed using this validated CFD model. Based on the CFD analyses, the final geometry of the gas entrainment mitigating devices has been selected and optimized. Finally the selected devices have been tested to confirm its performance for mitigation of gas entrainment in large scale models of reactor primary circuit and surge tank. This paper presents the studies carried out towards development of suitable gas entrainment mitigation devices for hot pool and surge tank of SFR. (author)

  8. Current status on fast reactor program in Kazakhstan

    International Nuclear Information System (INIS)

    Kazakhstan Atomic Scientific and Industrial Complex consists of uranium mining, fuel production, and power industry. On the territory of the former Semipalatinsk Nuclear Test Site, there are three research reactors (EWG-1M, thermal light water heterogeneous vessel reactor with light water moderator and coolant, beryllium reflector, maximum thermal power, 35 MW, 4 hours period of continuous operation at maximum power; IGR, impulse homogeneous uranium-graphite thermal reactor with graphite reflector, maximum heat release is 5.2 GJ (1 GJ in a pulse), maximum thermal neutron flux is 0.7*1017 cm-2s-1; RA, about 0.5 MW thermal high temperature heterogeneous reactor with air coolant, zirconium hydride moderator, and beryllium reflector), and one non-reactor test facility (EAGLE, reactor fuel element melting testing). One research reactor and sub-critical assembly near Almaty (VVR-K, 10 MW light water reactor) is used primarily for nuclear safety investigations. Following a Presidential decree, Kazakhstan will establish the following technology centres: Centre of Information Technologies, based at the Nuclear Physics Institute in Altau; Centre of Biotechnologies, based at the former military centre in Stepnogorsk; and the Centre of Nuclear Technologies, based at the National Nuclear Centre in Kurchatov City. The experimental reactor TOKOMAK will be constructed at Kurchatov City in support of the International Thermonuclear Experimental Reactor (ITER) project. Works have already started. The General Plan for the BN-350 decommissioning was developed within the framework of a Kazakh - US project. At the end of March 2003, the Plan was presented for final review to a IAEA group of experts. Due to a new US DOE initiative, of the Feasibility Study Report on the possibility to use 120 t metal-concrete casks for BN-350 spent fuel transportation and long-term storage was performed at the end of 2002. These casks shall be designed and manufactured in Russia. The content (NaK) of the

  9. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  10. A review of the UK fast reactor programme

    International Nuclear Information System (INIS)

    The fast reactor programme in the United Kindom is reviewed under the following headings: Progress with PFR; Reprocessing: Commercial Design Studies; Structural Integrity; Engineering and Components; Materials; Sodium Chemistry; Core and Fuel; Safety; Plant Performance. (author)

  11. Review of fast reactor activities at OECD (NEA)

    International Nuclear Information System (INIS)

    The Committee on the Safety of Nuclear Installations initiated several reports in 1979. Status reports are published on: the role of fission gas release in case of fuel element failure; reactivity monitoring in a LMFBR at shutdown; increasing the reliability of fast reactor shutdown systems. A report is planned on the interactions between sodium and concrete. LMFBR safety issue that were studied are concerned with containment R and D; natural circulation cooling; and fuel failure modelling. Nuclear Development Division was concerned with Gas cooled fast reactors technology. Nuclear Science Division dealt with fast reactor physics and nuclear data for fast reactors. NEA Data Bank provides technical support and acts as a computer code library and nuclear data centre

  12. Advances in sodium technology, testing and diagnostics of fast reactors

    International Nuclear Information System (INIS)

    The collection contains a selection of 29 papers from three international specialists' meetings: the CMEA conference ''Control and measuring instruments and diagnostic systems of fast reactors'' held in the GDR in April 1983; the IAEA conference on nuclear power experience held in Austria in September 1982; and the conference ''Problems of technology and corrosion in sodium coolant and protective gas'' held in the GDR in April 1977. Three papers on operating experience with Soviet fast reactors and their safety have a general character; they are followed up by three papers on sodium technology. Five papers deal with the diagnostics of fast sodium cooled reactors and nine papers are devoted to the diagnostics of steam generators. Eight papers relate to detectors for the diagnostics of fast reactors. Safety regulations for work with alkali metals are added. (A.K.)

  13. 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)

  14. Radioisotopes in the primary circuit of a fast reactor

    International Nuclear Information System (INIS)

    In the frame of the research performed to understand the behaviour of the radioactive isotopes of iodine in the primary coolant circuit of fast reactor, a simple theoretical model is proposed. Results concerning PHENIX and RAPSODIE are given

  15. A code to calculate multigroup constants for fast neutron reactor

    International Nuclear Information System (INIS)

    KQCS-2 code is a new improved version of KQCS code, which was designed to calculate multigroup constants for fast neutron reactor. The changes and improvements on KQCS are described in this paper. (author)

  16. Review of fast reactor operating experience gained in 1998 in Russia. General trends of future fast reactor development

    International Nuclear Information System (INIS)

    Review of the general state of nuclear power in Russia as for 1998 is given in brief in the paper. Results of operation of BR-10, BOR-60 and BN-600 fast reactors are presented as well as of scientific and technological escort of the BN-350 reactor. The paper outlines the current status and prospects of South-Urals and Beloyarskaya power unit projects with the BN-800 reactors. The main planned development trends on fast reactors are described concerning both new projects and R and D works. (author)

  17. Operating experience from the BN600 sodium fast reactor

    International Nuclear Information System (INIS)

    Conclusion: The operating experience from the BN600 reactor power unit for more than 32 years is positive in terms of the demonstration of the feasibility of the utilization of a sodium-cooled fast reactor for commercial electric generation. The BN600 reactor is an important key link ensuring the continuity and succession of the development of the fast reactors in Russia of which the reliable and steady operation confirms good prospects of this line of the nuclear power industry. In the course of the BN600 power unit operation the valuable operating experience from the individual systems and components which should be preserved and utilized when developing the advanced designs of the sodium-cooled fast reactors was accumulated

  18. New modelling method for fast reactor neutronic behaviours analysis

    International Nuclear Information System (INIS)

    Due to safety rules running on fourth generation reactors' core development, neutronics simulation tools have to be as accurate as never before. First part of this report enumerates every step of fast reactor's neutronics simulation implemented in current reference code: ECCO. Considering the field of fast reactors that meet criteria of fourth generation, ability of models to describe self-shielding phenomenon, to simulate neutrons leakage in a lattice of fuel assemblies and to produce representative macroscopic sections is evaluated. The second part of this thesis is dedicated to the simulation of fast reactors' core with steel reflector. These require the development of advanced methods of condensation and homogenization. Several methods are proposed and compared on a typical case: the ZONA2B core of MASURCA reactor. (author)

  19. Shuffling strategy study of breeding-burning integrated fast reactor

    International Nuclear Information System (INIS)

    The breeding-burning integrated fast reactor uses burning assemblies to generate thermal power, meanwhile, converts 238U into 239Pu in the fertile assemblies. With periodical shuffling of assemblies, the reactor can maintain criticality for decades of years. To maintain long-term stability of the core reactivity, the core layout and shuffling strategy should balance the burning and the breeding of the assemblies. The scattered core layout and shuffling strategy ensures fast breeding of the fertile assemblies, and keeps stable core power distribution in whole life of the reactor. Moreover, at the end of the reactor life, the discharge burnups of different fuel assemblies are close to each other, which are about 250300 GW · d/t. This is important for breeding-burning integrated fast reactor to achieve very efficient utilization of uranium resource without reprocessing. (authors)

  20. A review of fast reactor activities in Switzerland - March 1984

    International Nuclear Information System (INIS)

    As a result of the noncentralized government in Switzerland there is no clear national policy for the future application of nuclear energy. This is reflected in the lack of a generally agreed nuclear energy research policy in the country. Consequently, activities related to several advanced reactor concepts are funded simultaneously at similar, but relatively low levels. The total expenditure of 5.9 million Swiss Francs (approx. 1 SFr per capita) for fast reactor activities in 1983 must be judged in the light of this situation. The funds have been allocated to an LMFBR safety programme (52%) and a fuel development programme (48%). In the field of LMFBR safety analytical work is performed on hypothetical core disruptive accidents (HCDAs) and on the integrity of components under HCDA loadings with emphasis on the dynamic behaviour of the reactor cover. A considerable effort has recently been devoted to the preparations for the SONACO natural convection experiment. Another relatively new experimental activity, involving small-scale vapour explosions with freon and water, has produced evidence of interesting physical effects which are not in accord with the assumptions of current molten fuel-coolant interaction (MFCI) models. The fuel development programme has continued with the manufacture of spherepac mixed carbide fuel pins for an irradiation experiment in FFTF. However, the time scale of the experiment has suffered a set-back due to an accident in a glove box of the production line

  1. Recycle Strategies for Fast Reactors and Related Fuel Cycle Technologies

    International Nuclear Information System (INIS)

    Fast reactors and related fuel cycle (hereafter referred to as 'fast reactor cycle') technologies have the potential to contribute to long term energy security owing to their effective use of uranium and plutonium resources, and to a reduction in the heat generation and potential toxicity of high level radioactive wastes by burning long lived minor actinides recovered from spent fuel from light water reactors and fast reactors. Further, it is likely that fast reactor cycle technologies can play a certain role in non-proliferation as addressed in the Global Nuclear Energy Partnership. With these features, the research and development towards their commercialization has been promoted vigorously and globally as a future vision of nuclear energy. The introduction of fast reactor cycle systems will be carried out independently in each country according to its national conditions and nuclear energy policy. It should then be considered important to have a globally common consensus relating to safety philosophy, concepts of proliferation resistance, transuranic element burnup and recycling and so on. For the development and utilization of fast reactor cycle systems, while respecting each country's concept, it is essential to organize the technologies and concepts which countires should have in common globally and build a framework to make them standardized. The use of existing frameworks such as the Generation IV International Forum and the International Project on Innovative Nuclear Reactors and Fuel Cycles is considered effective to achieving this. Furthermore, a vigorous promotion such as international cooperative developments enables the formation of international consensus on major technologies for the fast reactor cycle as well as the saving of resources by infrastructure sharing. (author)

  2. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    OpenAIRE

    Lee, Seung Kyu; Kang, Byoung-Hwi; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spe...

  3. 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

  4. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  5. Improve Design of Fuel Shear for Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    GAO; Wei; OUYANG; Ying-gen; LI; Wei-min

    2012-01-01

    <正>Due to the deeper burnup and higher fuel swelling, fast reactor metal fuel rod using 316 stainless steel cladding, replacing the traditional zirconia cladding. The diameter of fuel rod of fast reactor is much longer than that of PWR, and the cladding of stainless steel has better ductility than zirconia cladding. Using the existing shear still will cause several aspects of problem: 1) Longer diameter of rod leads to

  6. Fast-neutron spectrum and absolute fast flux measurements in TR-1 reactor core and its reflector

    International Nuclear Information System (INIS)

    In this work, we tried to determine experimentally the parameter of an analytical expression for the fast neutron spectrum. We thus aimed to determine the fast neutron spectra in various location of the TR-1 (Cekmece Nuclear Research and Training Reactor). Threshold detectors In115, Ni58, AL27 were irradiated in different locations in the core and graphits regions of TR-1. Through the related activity measurements the parameters in question were found, thus the spectra could be calculated. The spectra is further used to compute the various spectrum avaraged neutron cross section in fast energy region

  7. 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.

  8. Status of Fusion Experimental Reactor (FER) design

    International Nuclear Information System (INIS)

    Conceptual design studies of the Fusion Experimental Reactor (FER) have been conducted at JAERI in line with a long-range plan for fusion reactor development laid out in the long-term program of the Atomic Energy Commission issued in 1982. The FER succeeding the tokamak device JT-60 is a tokamak reactor with a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. The paper describes recent developments of the FER design concept

  9. A review of calculation methods for fast and intermediate reactors

    International Nuclear Information System (INIS)

    This paper discusses the development of methods for calculating intermediate and fast reactors. It deals with various approaches to the problems of physical calculation. The calculation of resonance effects is discussed. Consideration is given to multi-group systems of fundamental and conjugate equations, various applications of perturbation theory to the problems of physical reactor calculation, and numerical methods of solving fundamental and conjugate reactor equations, which approximate the method of spherical harmonics. The paper describes an application of the response method to the solution of critical-mass problems, and methods of calculating reactors with hydrogeneous moderators. The fundamental features of an effective one-group reactor model are described. (author)

  10. High burnup fast reactor fuel: processing and waste management experiences

    International Nuclear Information System (INIS)

    The routine processing of mixed Plutonium/Uranium oxide fuels from the Prototype Fast Reactor (PFR) at Dounreay began in September 1980 and the design features of the modified Dounreay Fast Reactor (DFR) reprocessing plant and experience of the first active campaign were described in a paper to the British Nuclear Engineering Society in November 1981 (1). Since then progress in processing the fuel discharged from PFR has been covered briefly in a number of papers to international conferences and the Public Inquiry held in 1986 into the outline planning application for the proposed European Demonstration Reprocessing Plant. During this decade considerable experience in the operation of fast reactors and associated fuel plants has been accumulated providing confidence in the system before entering the next development phase - that of its commercial demonstration. Confidence in the UK draws on the successful operation of the PFR and the associated Dounreay fuel reprocessing and BNF Sellafield fabrication plants. Of equal importance is public confidence in safe operation and in the management of wastes generated by a fast reactor system. The present paper is a review of fast reactor reprocessing and waste management at the Dounreay Nuclear Establishment (DNE) as a contribution to the present status of the fast reactor system

  11. Fast reactor cover gas purification - The UK position

    International Nuclear Information System (INIS)

    The cover gas in the Prototype Fast Reactor (PFR) provides an inert gas blanket for both primary and secondary sodium circuits, ensures inert gas padding exists between the upper seals associated with penetrations through the reactor roof and provides argon to items of plant such as the control rods and the rotating shield and also to on line instruments such as the secondary circuit Katharometers. In order to meet these and other requirements purification of the argon cover gas is important to ensure: gas fed to purge gaps in the area of the magnetic hold device in the control rod mechanisms is not laden with sodium aerosols and reactive impurities (O2, H2) which could cause blocking both within the gaps and pipelines; gas phase detection systems which provide early warning of steam generator failures or oil ingress into the sodium are not affected by the presence of gaseous impurities such as H2, CO/CO2 and CH4; mass transfer processes involving both corrosion products and interstitial atoms cannot be sustained in the cover gas environment due to the presence of high levels of O2, N2 and carburising gases; background levels of radioactivity (eg Xe 133) are sufficiently low to enable gas phase detection of failed fuel pins, and the primary circuit gas blanket activity is sufficiently reduced so that discharges to the atmosphere are minimised. This paper describes how the PFR cover gas purification system is coping with these various items and how current thinking regarding the design of cover gas purification systems for a Civil Demonstration Fast Reactor (CDFR), where larger gas volumes and higher levels of radioactivity may be involved, is being guided by current experience on PFR. The paper also briefly review the experimental work planned to study aerosol and caesium behaviour in cove gas environments and discusses the behaviour of those impurities such as Zn, oil and N2 which are potentially damaging if certain levels are exceeded in operating plant

  12. Analysis of the fast transients in the research reactor MARIA

    International Nuclear Information System (INIS)

    The analysis of the physical process in the research reactor MARIA during fast reactivity excursions is presented. The mathematical model of dynamics has been developed and its parameters as well as the temperature coefficients of reactivity were analysed. The analysis has shown that the change in the coolant density and fuel expansion are the most dominant temperature effects of reactivity. These effects differ by nearly two orders of magnitude from the reactivity effect due to fuel temperature changes, while the influence of the remaining materials of the core, i.e. beryllium and graphite, is negligible. Weak fuel temperature feedback is caused chiefly by the high enrichment of U-235, which indicates that the role of the Doppler effect significant for fast transients is practically unimportant. On the basis of the dynamics model, the code TOTEM for the CYBER-73 digital computer was written in FORTRAN-EXTENDED. The user's manual with input preparation and output explanation is enclosed. The code is intended for analysis of the reactor accidents caused by the rapidly increased reactivity. The following calculations can be carried out: axial and radial temperature distribution in the fuel channel with mean power; feed-back reactivity; power transients. The code was verified experimentally. A comparison of the measured and calculated feed-back reactivity and temperatures has shown good agreement. The space effects were investigated using signals from two neutron detectors, cobalt self-powered detector (SPD-Co) and ionization chamber located at a different distance from the perturbation point. All considered accidents, excluding the stepwise insertion the reactivity of 2$ (on the assumption that the protection system would respond properly) do not peril the safety of the reactor. 43 refs., 39 figs., 7 tabs. (author)

  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. Performance of the lift-pump with the lead-bismuth cooled fast reactor. Experimental study on bubble distribution and circulation flow rate

    International Nuclear Information System (INIS)

    Recently, the utilization of the lift pump is examined in a small reactor of the lead-bismuth eutectic cooling. Then, the experiments concerning about void behavior and performance of the lift pump in three kinds of risers (1124mm in height and inside diameters φ69.3mm, φ106.3mm, φ155.2mm) were performed by using lead bismuth eutectic. The main results are as follows: (1) The local void fraction varies in horizontal plane in case of the big diameter riser. (2) The lead-bismuth circulating flow rate evaluated by a present design method becomes lower than that of experiments in case of medium and small diameter risers. This design method can be used as an outline evaluating function for these cases, considering the evaluation accuracy of the pressure loss of the test section in the calculation. (3) In the big diameter riser, the present design method excessively evaluates the lead-bismuth circulating flow rate. It thought that the circulation head will not occur in the experiments such as a results of the present design method because the void rises biasing in horizontal plane in case of big diameter riser though the present method is one dimensional model. It is better to utilize a separator which can divides the riser into about 10cm diameter flow path and the void is fed uniformly distributed to each paths to obtain appropriate circulation head. (author)

  15. Fast Pyrolysis of Lignin Using a Pyrolysis Centrifuge Reactor

    DEFF Research Database (Denmark)

    Trinh, Ngoc Trung; Jensen, Peter Arendt; Sárossy, Zsuzsa;

    2013-01-01

    Fast pyrolysis of lignin from an ethanol plant was investigated on a lab scale pyrolysis centrifuge reactor (PCR) with respect to pyrolysis temperature, reactor gas residence time, and feed rate. A maximal organic oil yield of 34 wt % dry basis (db) (bio-oil yield of 43 wt % db) is obtained at...

  16. Progress of Research on Demonstration Fast Reactor Main Pipe Material

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The main characteristics of the sodium pipe system in demonstration fast reactor are high-temperature, thin-wall and big-caliber, which is different from the high-pressure and thick-wall of the pressurized water reactor system, and the system is long-term

  17. Some policy aspects of the fast reactor question. Chapter 2

    International Nuclear Information System (INIS)

    The following aspects of energy policy in the UK are discussed: planning and forecasting, accuracy and relevance to government policies; economics; plant construction programmes; scope for electricity growth; arguments for and against fast reactor programme (and in relation to other types of reactor). The general discussions of energy policy cover coal, natural gas and oil in addition to nuclear power. (U.K.)

  18. CP ESFR: Collaborative Project for a European Sodium Fast Reactor

    International Nuclear Information System (INIS)

    The Collaborative Project for a European Sodium Fast Reactor (CP ESFR) is performed (2009-2012) in the 7th European Framework Programme. It is devoted to the identification and study of innovations to be considered for the future in the core design, safety, reactor architecture, components and the dissemination of knowledge related to this technology among young European professionals. (author)

  19. Knowledge Preservation and Data Collection on Fast Reactors in France

    International Nuclear Information System (INIS)

    Documentary packages: • MADONA - the French LMFR R&D Documentary collection; - A trilateral CEA, EDF & AREVA collection, 64 synthesis reports with 5300 relevant documents, on line. • RAPSODIE documentary collection - 18000 documents including 8000 drawings gathering available technical information on RAPSODIE; • PHENIX documentary fund - 140 000 documents. Design, specification, drawings, procedures, technical reports gathering available technical information on the PHENIX plant since 1968; • PHENIX lifetime extension project data base - 2000 reports. General methodology and feedback experience; • SUPERPHENIX Plant Archives - Several thousand of documents covering plant design, operation and decommissioning, since 1977; • SPX Safety report - 150 000 pages; - Plant characteristics & Technical options; - General safety analysis; - Manufacturing and operational rules (normal, incidental and anticipated operating conditions); • EFR 98 design documentation - around 5700 documents produced during the EFR studies; • EFR Synthesis reports - 48 synthesis reports; • ACCORE - 20 200 CEA technical documents on the period 1988-2001, with keyword research system; • The Fast Neutron Reactor System Knowledge Book - approx. 1400 licensing documents; • Sodium Technology data base - around 5500 technical documents on sodium circuits, sodium leak and fires, instrumentation, ISI&R, sodium water reaction in SG…; • CIR - The Reprocessing Information Centre data base gathered 25 000 reports, on reprocessing and analytical chemistry; among them around 600 reports concern Fast Reactor fuel reprocessing technology; • Database - Preservation of data relative to fast critical mock-ups: MASURCA (F), SNEAK(D), ZPPR (US), ZEBRA(UK), RRR/SEG(D). • ARCOPAC - Electronic data base on SUPERPHENIX physical measurements; • BREF Fuel Data Base - SFR fuels, behavior of structural materials; • Fatigue experimental data base for 316LN; • PHENIX expert Data base - Built

  20. Advanced sodium fast reactor accident source terms : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  1. PROTEUS investigations for advanced thermal, fast, and intermediate-spectrum reactors

    International Nuclear Information System (INIS)

    The zero-power reactor, PROTEUS, has been used over the years for physics investigations concerning various types of advanced systems, namely gas-cooled fast reactors (GCFRs) in the seventies, light water high-conversion reactors (LWHCR) in the eighties and, currently, low-enriched-uranium high-temperature reactors (LEU-HTRs). The wide range of test neutron spectra cover underlines the versatility of the facility, the safety and operational limits for which have largely remained unaltered during the different experimental programs. This paper reviews the scope of the various PROTEUS investigations and includes evaluations of some of the most recent experiments

  2. Technical Meeting on Existing and Proposed Experimental Facilities for Fast Neutron Systems. Working Material

    International Nuclear Information System (INIS)

    The objective of the TM on “Existing and proposed experimental facilities for fast neutron systems” was threefold: 1) presenting and exchanging information about existing and planned experimental facilities in support of the development of innovative fast neutron systems; 2) allow creating a catalogue of existing and planned experimental facilities currently operated/developed within national or international fast reactors programmes; 3) once a clear picture of the existing experimental infrastructures is defined, new experimental facilities are discussed and proposed, on the basis of the identified R&D needs

  3. Future experimental programmes in the CROCUS reactor

    OpenAIRE

    Lamirand, Vincent Pierre; Hursin, Mathieu; Perret, Grégory; Frajtag, Pavel; Pakari, Oskari; Pautz, Andreas

    2016-01-01

    CROCUS is a teaching and research zero-power reactor operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology (EPFL). Three new experimental programmes are scheduled for the forthcoming years. The first programme consists in an experimental investigation of mechanical noise induced by fuel rods vibrations. An in-core device has been designed for allowing the displacement of up to 18 uranium metal fuel rods in the core periphery. ...

  4. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

  5. Fast reactors: the future of nuclear energy

    International Nuclear Information System (INIS)

    The main problems to be solved for FBR type reactors become viable economically, presenting the research programs of Europe, United States of America, Japan and Brazil are described. The cooperations between interested countries for improving FBR type reactors, and the financial and human resources necessaries for the development of programs, are evaluated. The fuel cycle is also analysed. (M.C.K.)

  6. Status of national programmes on fast reactors in Korea

    International Nuclear Information System (INIS)

    The role of nuclear power plants in electricity generation in Korea is expected to become more important in the years to come due to poor natural resources and green house gases. This heavy dependence on nuclear power eventually raises the issues of efficient utilization of uranium resources and of spent fuel storage. Fast reactors can resolve these issues. Korea Atomic Energy Research Institute started development of a Liquid Metal Reactor design in 1997 and completed the Conceptual Design in March of 2002. Efforts are currently directed toward the development of advanced fast reactor concepts and basic key technologies. (author)

  7. Review of the United Kingdom fast reactor programme - March 1986

    International Nuclear Information System (INIS)

    The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (2) progress with the prototype fast reactor (PFR) which achieved its design power on 4 March 1985; (3) nuclear fuel reprocessing; (4) commercial design studies; (5) structural integrity of LMFBR during its lifetime; (6) R and D work on components of LMFBR; (7) materials study; (8) sodium chemistry; (9) reactor core and fuel design philosophy; (10) safety problems; (11) plant performance studies

  8. Method and data investigations for Pu-burning fast reactor configurations

    International Nuclear Information System (INIS)

    Fast reactors have a potentially important role to play in the future as they effectively burn plutonium and minor actinides. However, the core characteristics of plutonium- and actinide-burning fast reactors provide new challenges to existing methods and data. Various numerical and experimental investigations have therefore been undertaken to assess the validity of the CEA core calculation schemes applied to Pu-burning fast reactors. The deterministic scheme of calculations based on the JEF-2 evaluations and on the ECCO/ERANOS code system has been compared with independent evaluations based on other code systems and on the Monte Carlo method. The ECCO/ERANOS predictions have also been compared with differential and integral measurements provided by the ongoing CIRANO experimental programme in the MASURCA facility. Preliminary results of these validation and qualification efforts are presented in this paper. The directions of future collaborative work between PSI and CEA is also indicated. (author) 11 figs., 6 tabs., 6 refs

  9. The case for the gas cooled fast reactor

    International Nuclear Information System (INIS)

    Although gas-cooling for fast reactors had been the subject of consideration since the early days of nuclear power, it was when the concept of the prestressed concrete pressure vessel turned into practical fact, that convincing arguments could be made to overcome safety objections. In terms of hardware, the Gas Cooled Fast Breeder Reactor can rely on existing and available technologies; as far as fuel is concerned, valuable information will be derived from the Liquid Metal Fast Breeder Reactor programme. The GCFR can be made very flexible; its capital cost will not exceed by more than 20% the one for reactor built at present on commercial scale; the overall economy of its fuel cycle is good. It could play an important role in the future breeder family

  10. Fast-power-reactor optimization by the game theory

    International Nuclear Information System (INIS)

    In the first stage of the use of fast breeder reactor - because fissile-material amounts are small - we are interested in fast breeder reactors which achieve minimum fissile-material mass, with maximum power. This problem shows a two-matrix-game structure. First, we determine a competive-game solution and second, a cooperative-game solution, obtaining in this way the optimum distribution of the fissile and fertile materials in the multizone fast reactors. Another optimization problem which is solved in this paper is finding the reactor structure for which the power non-uniformity factor and the flux non-uniformity factor are minimum. This is, also, a mathematical two-matrix game and it is solved as above. The two optimization problems have different solutions. (author)

  11. Fast neutron reactors: a long experience facing a new future

    International Nuclear Information System (INIS)

    This article makes a survey of the different fast reactor programs throughout the world. The European fast reactor project (EFR) was launched in 1988 under the impulse of a partnership involving electricity producers, nuclear core manufacturers and research agencies. The fading prospect of an energy shortage has led to the freeze of EFR project by the end of 1998. In Europe most fast reactor programs have entered a waiting period whose duration could reach decades. Nevertheless the necessity of assuring the capability of designing and building such reactors stays a priority in order to benefit from the property of breeding which could strongly contribute to the future energetic autonomy of Europe. The improvement of technical performances, the integration of technological progress, the investigation of new concepts are the main tasks of the waiting period. Some prototypes will have to be built at regular time intervals in order to assure the feasibility of these evolutions. (A.C.)

  12. Simulator platform for fast reactor operation and safety technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  13. Research on the usage of a deep sea fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-09-01

    Many new types of fast reactors have been studied in PNC. A deep sea fast reactor has the highest realization probability of the reactors studied because its development is desired by many specialists of oceanography, meteorology, deep sea bottom oil field, seismology and so on and because the development does not cost big budget and few technical problems remain to be solved. This report explains the outline and the usage of the reactor of 40 kWe and 200 to 400 kWe. The reactor can be used as a power source at an unmanned base for long term climate prediction and the earth science and an oil production base in a deep sea region. On the other hand, it is used for heat and electric power supply to a laboratory in the polar region. In future, it will be used in the space. At the present time, a large FBR development plan does not proceed successfully and a realization goal time of FBR has gone later and later. We think that it is the most important to develop the reactor as fast as possible and to plant a fast reactor technique in our present society. (author)

  14. Research and development studies carried out for the seismic verification of the Italian PEC Fast Reactor

    International Nuclear Information System (INIS)

    This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA, in co-operation with ANSALDO and ISMES for the seismic verification of the Italian PEC Fast Reactor. More precisely, the paper focuses on the wide-ranging research and development programme that was performed on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the general validity of the analyses in the framework of research and development activities for nuclear reactors are pointed out. The adopted design criteria and methods are presented in a separate paper, together with the effects of seismic conditions on PEC design, and comparisons with the other fast reactors of the European Community countries. (author)

  15. Path to a commercial fast reactor option in the United States

    International Nuclear Information System (INIS)

    Fast reactors represent one technology with the potential to transform the energy sector through increased resource utilization and unique waste management opportunities that are not available through the established light water reactor (LWR) technology. The technology is 'old' by 21 st century standards, as the United States and other nations have funded fast reactor research and development efforts from the birth of nuclear power that have led to deployment at all scales from experimental to commercial. Yet, in spite of great promise, the experience with fast reactors can be characterized as mixed, with both notable successes and failures. Why then, after 50 years of development, is there no commercial fast reactor operating in the United States? Is commercialization of the technology possible? If so, what business case is needed for a commercial utility to operate a fast reactor? This paper will attempt to sketch a possible path for the development and demonstration of technology needed to make fast reactors a real option for commercial operation by mid-century. This path starts with today's proven LWR technology and uranium oxide fuels and takes an evolutionary route to a first commercial fast reactor that complements and supports the existing reliable LWR fleet. This paper is not intended to promote any one technology or approach, but instead seeks to illustrate the value of national level support for an energy technology program oriented toward demonstration and seeks eventual deployment of transformational energy systems on the scale and time-frame that matters. The observations and conclusions presented are derived primarily from the collective experience and expertise of the authors who represent a broad range of commercial perspectives from the utility, vendor, research and development, and regulatory policy communities. (authors)

  16. Fast Reactor Programme. Second Quarter 1969. Progress Report. RCN Report

    International Nuclear Information System (INIS)

    This progress report covers fast reactor research carried out by RCN during the second quarter 1969 forming part of the integrated fast breeder research and development programme also in progress at the national nuclear research centres Karlsruhe and Mol. The combined effort is based on a memorandum of co-operation in this field signed by the respective governments in 1968 and on a memorandum of understanding signed by the research centres. The RCN contribution is mainly concerned with the core of the fast breeder reactor and related safety aspects and, as such, must be looked upon as being complementary to the industrial research pro- field of fast reactors. The contribution comprises the following six items: - A Æéatîtôr , physics programme to determine the influence of fission products on several main characteristics of the reactor core such as void coefficient, Doppler coefficient and breeding ratio; - A fuel performance programme in which both stationary and transient irradiations are being carried out to establish the temperature and power limits of fuel rods; also the consequences of loss- of-cooling will be investigated; - Investigation into the change in mechanical properties of fuel canning materials due to high fast neutron doses; - A study of the corrosion behaviour of canning materials and their compatibility with the fuel under conditions of high temperature and high pressure; - Investigation into the behaviour of aerosols of fission products which could be formed after a fast reactor accident; a thorough understanding is of utmost importance for the reactor safety assessment ; - Studies on heat transfer in the reactor core. As fast breeders operate at high power densities, an accurate knowledge on the heat transfer phenomena is required

  17. OECD Nuclear Energy Agency Activities Related to Fast Reactor Development

    International Nuclear Information System (INIS)

    The OECD Nuclear Energy Agency (NEA), whose role is to assist its member countries to develop, through international cooperation, the scientific and technological bases required for the safe, environmentally friendly and economical use of nuclear energy, conducts work related to fast reactor systems in two areas of activity: one focused on scientific research and technology development needs and one dedicated to strategic and policy issues. Recent, scientifically oriented, fast reactor related activities coordinated by the NEA comprise: -A coordinated effort to evaluate basic nuclear data needed for the development of fast reactor systems; -A recently initiated review of Integral Experiments for Minor Actinide Management; -An ongoing study on Homogeneous versus Heterogeneous Recycle of Transuranic Isotopes in Fast Reactors; -A comparative analysis of the safety characteristics of sodium cooled fast reactors; -A series of workshops on Advanced Reactors with Innovative Fuels; -A series of information exchange meetings on actinide and fission product partitioning and transmutation. The NEA has also conducted two reviews on issues related to the transition from thermal to fast neutron nuclear systems. One study was devoted to technical issues, including benchmark studies on: (i) the performance of scenario analysis codes, (ii) a regional (European) scenario and (iii) a global transition scenario. The other study emphasized issues of interest to policymakers, such as key parameters affecting the cost-benefit analysis of transitioning, including the size and age of the nuclear reactor fleet, the expected future reliance on nuclear energy, access to uranium resources, domestic nuclear infrastructure and technology development, and radioactive waste management policy in place. The NEA is also an active player in many other international activities related to fast neutron systems, such as the Generation IV International Forum, where the NEA acts as technical secretariat for

  18. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  19. Assumed mode approach to fast reactor core seismic analysis

    International Nuclear Information System (INIS)

    The need for a time history approach, rather than a response spectrum approach, to the seismic analysis of fast breeder reactor core structures is described. The use of a Rayleigh-Ritz/Assumed Mode formalism for developing mathematical models of reactor cores is presented. Various factors including structural nonlinearity, fluid inertia, and impact which necessitate abandonment of response spectrum methods are discussed. The use of the assumed mode formalism is described in some detail as it applies to reactor core seismic analysis. To illustrate the use of this formal approach to mathematical modeling, a sample reactor problem with increasing complexities of modeling is presented. Finally, several problem areas--fluid inertia, fluid damping, coulomb friction, impact, and modal choice--are discussed with emphasis on research needs for use in fast reactor seismic analysis

  20. A Review of the UK Fast Reactor Programme: March 1980

    International Nuclear Information System (INIS)

    Towards the end of 1979 the Government announced a new programme of thermal reactor stations to be built over ten years (totalling 15GW), in addition to the two AGR stations at Torness and Heysham 'B' which had been approved by the previous Government. The first station of the new programme will be based on a Westinghouse PWR, subject to safety clearance and the outcome of a public inquiry, and it is envisaged that the remaining stations of the programme would be split between PWRs and AGRs. The AEA Chairman wrote formally to the Secretary of State for Energy in December 1979, putting forward on behalf of the Electricity Supply Authorities, NNC, BNFL and the AEA a recommended strategy for building the Commercial Demonstration Fast Reactor (CDFR), subject to normal licensing procedure and to public inquiry, so as to ensure that the key options for introducing commercial fast reactors, when required, should remain open. A Government statement is expected during the next few months. Meanwhile the level of effort on fast reactor research and development in the UK has been maintained, the fast reactor remaining the largest of the UKAEA's reactor development projects with expenditure totalling somewhat over £80M per annum. The main feature of the UK fast reactor programme has continued to be the operation of PFR (Sections 2 and 7) which is yielding a wealth of experience and of information relevant to the design of commercial fast reactors. Bum-up of standard driver fuel has reached 6-7% by heavy atoms, while specially enriched lead fuel pins have reached 11 % without failure. An extensive programme of work in the reactor and its associated steam plant was completed in March 1980 and the reactor then started its fifth power run. The fuel reprocessing plant at DNE is being commissioned and has reprocessed some of the spent fuel remaining from the DFR. It will start soon on reprocessing fuel discharged from the PFR. During the year improvements to the design of the future

  1. 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)

  2. Design features of BREST reactors and experimental work to advance the concept of BREST reactors

    International Nuclear Information System (INIS)

    Principle design features of BREST-300 (300 MWe) and BREST-1200 (1200 MWe) lead.cooled fast reactors are presented in this paper. Several experimental works have been performed or under way in order to justify lead-cooled reactor design concepts. BREST reactor designs of different outputs have been developed using the same principles. In conjunction with the increased output and the implement of inherent safety concept, a number of new solutions, which may be applied to the BREST-300 reactor design too, have been considered in the BREST-1200 reactor design. The new design features adopted in the BREST-1200 reactor design include: pool-type reactor design not requiring metal vessel, hence, not limiting reactor power; new handling system allowing to reduce central hall and building dimensions as a whole; emergency cooling system using field pipes, immersed directly in lead, which may be used to cool down reactor under normal conditions; by--pass line incorporated in coolant loop allowing to refuse the actively actuating valve initiated in pumps shut down. (author)

  3. 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

  4. Education and Training in Support of Sodium Cooled Fast Reactors Around the World

    International Nuclear Information System (INIS)

    The Generation IV Technology Roadmap has identified six systems for their potential to meet the new technology goals to improve safety, sustainability, economic competitiveness and proliferation resistance. Among these systems, three are fast neutron reactors: two cooled by liquid metal, the sodium cooled fast reactor (SFR) and the lead cooled fast reactor, and one cooled by gas, the gas cooled fast reactor. The SFR has the most comprehensive technological basis as a result of the experience gained from worldwide operation of several experimental, prototype and commercial size reactors from the 1940s. In order to support the operation of existing reactors, design activities for new projects and decommissioning of old reactors, it is mandatory to maintain and develop skills, particularly among the young generation. This paper presents the current strategies developed at the national level, or within a multilateral framework such as the EU or the IAEA, to support the development of SFRs, with particular focus on education and training initiatives dedicated to students, researchers, designers and operators involved in the development of SFRs. (author)

  5. Current status on fast reactor program in Kazakhstan

    International Nuclear Information System (INIS)

    Atomic scientific-industrial complex of Kazakhstan deals with uranium mining, production and power industry. The three types of research reactors existing are used for research mostly in the field reactor safety. The BN-350 reactor is being decommissioned within the frame of Kazakhstan-American project reviewed by IAEA. The EAGLE project is carried out dealing with design of an extremely safe fast reactor. Kazakstan is taking part in IAEA CRP on 'Fission product yield data required for transmutation of minor actinide nuclear waste'

  6. Status of fast reactor activities in the Russian Federation

    International Nuclear Information System (INIS)

    The power production program was developed before the disintegration of the USSR and CIS. This report covers therefore the current status of power production and consumption in in republics of the former USSR with a separate chapter on the status of nuclear power. It covers some general results concerned with fast reactors operational experience and BN-600 power plant operational experience. This includes radiological conditions at the BN-600 and reactor core operating experience. Separate chapters are devoted to BN-350, BOR-60, BR-10 and BN-800 reactors. Work devoted to large-size reactor design are described including research and development and fabrication

  7. Towards the thorium fuel cycle with molten salt fast reactors

    International Nuclear Information System (INIS)

    Highlights: • Neutronic calculations for fast spectrum molten salt reactor. • Evaluation of the fissile matter to be used in such reactor as initial fissile load. • Capabilities to transmute transuranic elements. • Deployment scenarios of the Thorium fuel cycle. • Waste management optimization with molten salt fast reactor. - Abstract: There is currently a renewed interest in molten salt reactors, due to recent conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts. It has been recognized as a long term alternative to solid-fueled fast neutron systems with a unique potential (large negative temperature and void coefficients, lower fissile inventory, no initial criticality reserve, simplified fuel cycle, wastes reduction etc.) and is thus one of the reference reactors of the Generation IV International Forum. In the MSFR, the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this characteristic, the MSFR can operate with widely varying fuel compositions, so that the MSFR concept may use as initial fissile load, 233U or enriched uranium or also the transuranic elements currently produced by light water reactors. This paper addresses the characteristics of these different launching modes of the MSFR and the Thorium fuel cycle, in terms of safety, proliferation, breeding, and deployment capacities of these reactor configurations. To illustrate the deployment capacities of the MSFR concept, a French nuclear deployment scenario is finally presented, demonstrating that launching the Thorium fuel cycle is easily feasible while closing the current fuel cycle and optimizing the long-term waste management via stockpile incineration in MSRs

  8. Irradiation Experiments on Plutonium Fuels for Fast Reactors

    International Nuclear Information System (INIS)

    An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O2 -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650°C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO2 and(U,Pu)O2 particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to ∼ 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent experiment was conducted in a thermal

  9. 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

  10. Analysis of deficiencies in fast reactor blanket physics predictions

    International Nuclear Information System (INIS)

    This analysis addresses a deviation between experimental measurements and fast reactor blanket physics predictions. A review of worldwide results reveals that reaction rates in the blanket are underpredicted with the discrepancy increasing with penetration into the blanket. The analysis of this discrepancy involves two parts: quantifying possible error reductions using the most advanced methods and investigating deficiencies in current methodology. The source of these discrepancies was investigated by application of ''state-of-the-art'' group constant generation and flux prediction methodology to flux calculations for the Purdue University Fast Breeder Blanket Facility (FBBF). Refined group constant generation methods yielded a significant reduction in the blanket deviations; however, only about half of the discrepancy can be accounted for in this manner. Transport theory calculations were used to predict the blanket neutron transmission problem. The surprising result is that transport theory predictions utilizing diffusion theory group constants did not improve the blanket results. Transport theory predictions exhibited blanket underpredictions similar to the diffusion theory results. The residual blanket discrepancies not explained using advanced methods require a refinement of the theory. For this purpose an analysis of deficiencies in current methodology was performed

  11. Cross section weighting spectrum for fast reactor analysis

    International Nuclear Information System (INIS)

    Preparation of a nuclear data library is the first task that a reactor analyst needs to perform a neutronic analysis of a reactor type. Today, in fast reactor area, the scheme used to generate this library includes the processing of an evaluated nuclear data file to obtain cross sections, in thousands of groups. Sequentially, the nuclear data are processed by a cell code to obtain neutron flux that is used to condense the large amount of energy groups to a practical calculation number of groups that can be used in reactor analysis. In the first step of the scheme it is necessary a weighting spectrum to generate the nuclear data. Here, it is proposed to use the flux estimated by Monte Carlo code using cell or the exact geometries and actual composition of the problem to obtain the main portion of the weighting spectrum instead of a code built-in function. As an example, it is presented the differences between selected pins spectrums obtained with MCNP5 calculation of a hexagonal fast reactor fuel assembly. Also, it is showed a comparison between these spectra and the one obtained in the representative unit-cell model of this fuel assembly. The comparisons support that the proposed procedure, problem dependent, may be more accurate and a good choice to generate weighting spectrum in ultra-fine energy structure for fast reactor analysis. The proposed method will be used in space reactor neutronic analysis. (author)

  12. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

  13. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in a previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes were considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the ''in-situ'' replacement of first walls using atomic coating processes were considered. The vapor deposition of carbon was shown to be promising

  14. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in the previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes was considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the in-situ replacement of first walls using atomic coating processes was considered. The vapor deposition of carbon was shown to be promising

  15. Maintenance equipment for a fast reactor

    International Nuclear Information System (INIS)

    Object: To permit cleaning of through-holes of various diameters by mounting a spray nozzle on the lower end of a vertically movable extension tube connected to an Ar gas heater. Structure: After removing control rod drive mechanism and other apparatus mounted on the top of the reactor core for maintenance, a spray nozzle of a maintenance apparatus on a reactor top pit lid is inserted into a through-hole in a shield plug. Then, Ar gas heated by the heater is supplied through the extension tube and sprayed, thereby removing Na slug attached to and solidified on the inner surface of the hole by fusion. (Seki, T.)

  16. Preliminary evaluation of alternate-fueled gas cooled fast reactors

    International Nuclear Information System (INIS)

    A preliminary evaluation of various alternative fuel cycles for the Gas-Cooled Fast Reactor (GCFR) is presented. Both homogeneous and heterogeneous oxide-fueled GCFRs are considered. The scenario considered is the energy center/dispersed reactor concept in which proliferation-resistant denatured reactors are coupled to 233U production reactors operating in secure energy centers. Individual reactor performance characteristics and symbiotic system parameters are summarized for several possible alternative fuel concepts. Comparisons are made between the classical homogeneous GCFR and the advanced heterogeneous concept on the basis of breeding ratio, doubling time, and net fissile gain. In addition, comparisons are made between a three-dimensional reactor model and the R-Z heterogeneous configuration utilized for the depletion and fuel management calculations. Lastly, thirty-year mass balance data are given for the various GCFR fuel cycles studied

  17. Structural materials for Russian fast reactor cores. Status and prospects

    International Nuclear Information System (INIS)

    Full text: The energy strategy of Russia in the period up to 2020 contemplates a gradual introduction of a new nuclear energy technology based on the fast breeder reactors with the closed MOX fuel cycle. Further developments of nuclear power will demand inclusion of fast breeder reactors into the structure of NPPs. Since 1980 in Russia at Beloyarsk NPP the only in the word commercial fast breeder reactor BN-600 is in operation. According to the plans the fourth power unit at Beloyarsk NPP with the fast breeder reactor BN-800 shall be put into operation in 2012. Under developments is a commercial sodium cooled fast breeder BN-1800. The use of steel EP450 (12Cr13Mo2NbVB) as shrouds of FAs (96x2 mm) and cold worked steel ChS68 (06Cr16Ni15Mo2Mn2TiVB) as fuel claddings (6,9x0,4 mm) reliably ensured the fail-free operation of BN-600 reactor at the burnup of 11.2 % h.a. and the damage dose of 82 dpa. There is every reason to assume that the EP450 steel shrouds will not limit to reaching a higher fuel burnup. Currently, for the BN-type reactors as promising structural materials for a staged increase in the fuel burn-up under consideration are austenitic and martensitic steels including those produced by the powder metallurgy method (ODS steels). The main cause that restricts the burn-up of fuel clad in austenitic steels is their considerable swelling. This fact in its turn is responsible for the degradation of cladding short-time and long-time mechanical properties. Consideration has been given to the principles of complex alloying and treatment of austenitic steels that make low swelling feasible at the irradiation doses of ∼ 100 dpa. Currently experiments are under way in BN-600 reactor to validate the serviceability of austenitic steels as claddings: ChS68 steel up to ∼ 90 dpa, EK164 steel (07Cr16Ni19Mo2Mn2TiVB) up to ∼ 100 dpa. As a cladding material that provides for the fuel rod operation to the damage doses of ∼140 dpa under consideration are high

  18. Fast Development Of China's Small Satellite Industry

    Institute of Scientific and Technical Information of China (English)

    Sun Hongjin

    2009-01-01

    @@ China Spacesat Co., Ltd of China Academy of Space Technology (CAST) recently said, along with the successful launch of HJ-1A/B for the environment and disaster monitoring and forecasting small satellite constellation and after years of efforts, small satellite development technology has achieved fruitful results, and the development status has been greatly improved.China's small satellite technology has realized a great-leap-forward in development from a single satellite model to series model, from the satellite program to space industry. China has explored a development road for China's small satellite industrialization, and a modern small satellite development base has resulted.

  19. Education & Training in Support to Sodium Fast Reactors Around the World

    International Nuclear Information System (INIS)

    The results of these ambitious and long term strategies are: - first the creation of a new generation of skilled nuclear engineers in the field. - secondly a share of knowledge gained through experimental studies carried out in research laboratories as well as feedback from fast reactors operation, - thirdly a standardized information on safety, - and finally the creation of a “Sodium Fast Reactor community” is promoted, able to debate, share the knowledge and suggest new tracks for a better definition of design and operating rules

  20. Fuel damage during off-normal transients in metal-fueled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, J.M.; Bauer, T.H.

    1990-01-01

    Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs.

  1. Fast ultrasonic visualisation under sodium. Application to the fast neutron reactors

    International Nuclear Information System (INIS)

    The fast ultrasonic visualization under sodium is in the programme of research and development on the inspection inside the fast neutron reactors. This work is about the development of a such system of fast ultrasonic imaging under sodium, in order to improve the existing visualization systems. This system is based on the principle of orthogonal imaging, it uses two linear antennas with an important dephasing having 128 piezo-composite elements of central frequency equal to 1.6 MHz. (N.C.)

  2. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    International Nuclear Information System (INIS)

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  3. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

  4. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, Stephen Rhead [Idaho National Lab. (INL), Idaho Falls, ID (United States); Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  5. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

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

  6. Construction of Soviet fast reactor BN-600

    International Nuclear Information System (INIS)

    A sectional view is shown of the integral configuration of the 3rd unit reactor in the Beloyarsk nuclear power plant. The reactor vessel is a cylinder 12.8 m in diameter and 12.6 m in height. In view of overpressure in the vessel (40 kPa) the wall thickness is 30 to 40 mm. The reactor core contains 370 hexagonal fuel elements. Each element consists of 127 pins of an outer diameter of 6.9 mm. 27 positions are taken by regulating and scram rods. The fuel reserve in the core and the efficiency of reactivity control permits reactor operation for about 150 days such that one third of the fuel elements is exchanged during refuelling. A block diagram is shown of the power plant heat generating system. Core cooling is ensured by three circuits, i.e., the sodium primary and secondary circuits and one water and steam circuit. The progress of the power plant construction is briefly indicated. (J.P.)

  7. Capital cost: gas cooled fast reactor plant

    International Nuclear Information System (INIS)

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

  8. The seismic assessment of fast reactor cores in the UK

    International Nuclear Information System (INIS)

    The design of the UK Commercial Demonstration Fast Reactor (CDFR) has evolved over a number of years. The design has to meet two seismic requirements: (i) the reactor must cause no hazard to the public during or after the Safe Shutdown Earthquake (SSE); (ii) there must be no sudden reduction in safety for an earthquake exceeding the SSE. The core is a complicated component in the whole reactor. It is usually represented in a very simplified manner in the seismic assessment of the whole reactor station. From this calculation, a time history or response spectrum can be generated for the diagrid, which supports the core, and for the above core structure, which supports the main absorber rods. These data may then be used to perform a detailed assessment of the reactor core. A new simplified model of the core response may then be made and used in a further calculation of the whole reactor. The calculation of the core response only, is considered in the remainder of this paper. One important feature of the fast reactor core, compared with other reactors, is that the components are relatively thin and flexible to promote neutron economy and heat transfer. A further important feature is that there are very small gaps between the wrapper tubes. This leads to very strong fluid-coupling effects. These effects are likely to be beneficial, but adequate techniques to calculate them are only just being developed. 9 refs, figs

  9. Small size modular fast reactors in large scale nuclear power

    International Nuclear Information System (INIS)

    The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

  10. Advanced monitoring and control systems for fast reactors

    International Nuclear Information System (INIS)

    One of the important aspects of nuclear power station (NPS) improvement with fast reactors is provision of safety. The safety conception of advanced fast power reactors is directed on elaborating such solutions where as much as possible properties of reactor self-protection and natural laws are used in which the self-protection of the nuclear reactor is realized. To these solutions we may refer the usage of hydraulically weighted rods of alarm protection, negative temperature and power coefficients, negative sodium empty effect, natural circulation without power sources, natural convection and other measures. Additionally special technological systems are envisaged, which start functioning with the coming of the initial event of the accident. 1 ref., 7 figs, 1 tab

  11. A review of fast reactor program in Japan. April 2000 - March 2001

    International Nuclear Information System (INIS)

    This report describes the development and activities on fast reactors in Japan thru April 2000 to March 2001. During this period, the most important result of the Japanese Fast Reactor Project was that the first phase 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was completed at the end of March 2001, and the second phase study has just started in order to narrow down the candidate concepts selected in the first phase for next stage. In the Experimental Fast Reactor 'Joyo', the 35th rated power operation was completed by the end of May 2000. The 13th periodical inspection and reconstruction works for the Joyo upgrading program (MK-III) were started on the beginning of June 2000. The modification of the cooling system is underway. In the Prototype Fast Breeder Reactor 'Monju', countermeasures against sodium leakage have already been drawn up based on 'Monju' comprehensive safety review. The Japan Atomic Energy Commission (JAEC) has issued a new 'Long-term Program for Research, Development and Utilization of Nuclear Energy' in November 2000. (author)

  12. A review of fast reactor program in Japan. April 1999 - March 2000

    International Nuclear Information System (INIS)

    This report describes the development and activities of fast reactors in Japan through April 1999 to March 2000. During this period, the most important result of the Japanese Fast Breeder Reactor (FBR) Project was the phase-1 feasibility studies on commercialized fast reactor cycle system which started in July 1999, associating with electric power companies. In the Experimental Fast Reactor 'Joyo', the 33rd and 34th rated power operation carried out after the 12th periodical inspection. The 35th rated power operation is under continuance by the end of May 2000. The 13th periodical inspection and reconstruction works for the Joyo upgrading program (MK-III) will be started at the beginning of June 2000. In the Prototype Fast Breeder Reactor 'Monju', countermeasures against sodium leakage are being examined according to the Monju comprehensive safety review. The 'Long-term Program for Development and Utilization of Nuclear Energy' is now under deliberation in the special committee of the Japan Atomic Energy Commission. The Commission will issue the new plan by the end of 2000. (author)

  13. Sodium components cleaning status in the Italian fast reactor program

    International Nuclear Information System (INIS)

    As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed

  14. Studies on the transient operation and stability of fast reactors

    International Nuclear Information System (INIS)

    These studies form part of the general programme of perfecting calculation methods for fast reactors. The basic formulae are given for the layouts used, i.e. the classic kinetic and thermal exchange equations, etc. A description is then given of the digital computer methods employed for studying the stable functioning of the reactor and of the methods used for transient operation studies. Finally, some examples of application are discussed and a comparison is made with parallel studies on the same subject. (author)

  15. Proceedings of 'workshop on Pb-alloy cooled fast reactor'

    International Nuclear Information System (INIS)

    The objective of 'Workshop on Pb-Alloy Cooled Fast Reactor', held in Taejeon, Korea on May 6, 2003, is to enhance the basic knowledge in this area by facilitating the exchange of information and discussions about problematic area of design aspects. There were five presentations from three different countries and about 25 participants gathered during the workshop. The topics covered in the workshop include benefits and drawbacks of Pb-alloy and Sodium coolant, two Pb-alloy cooled 900 MWt reactor designs using both B4C rods and NSTs, BREST-300 breakeven reactor and transmutation effectiveness of LLFPs in the typical thermal/fast neutron systems. The generic conclusion for the Pb-alloy cooled fast reactor from this workshop is as follows: 1) It has a potential to satisfy the goals established for the Generation-IV reactor concepts, so it has a bright future. 2) As a fast neutron system with a moderate breeding or a conversion, it is flexible in its roles and has superior safety characteristics over sodium coolant because of Pb-alloy's chemical inertness with water/air and high boiling temperature

  16. Fast reactor 3D core and burnup analysis using VESTA

    Energy Technology Data Exchange (ETDEWEB)

    Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

  17. UK fast reactor components - sodium removal decontamination and requalification

    International Nuclear Information System (INIS)

    Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field

  18. Opening Address: Fast Neutron Reactors and Sustainable Development

    International Nuclear Information System (INIS)

    The aim of this presentation is to provide an insight into the challenges that lie ahead for the development of fast reactors. From the moment when the first fast reactor - EBR1 - lit up the city of Arco right up to Superphenix, by far the largest fast reactor ever built, there have been 40 years of fast reactor development, mainly centred on sodium cooled systems, leading to the successful operation of such plants. Therefore, the question could arise about the need for more R and D and the relevance of new prototype designs. There have been two major development steps in the history of fast reactors. During the 1960s and 1970s, their development was undertaken following concerns related to the energy supply, resulting mainly from the oil crisis, as well as from the need to use uranium resources more efficiently. In the 1980s, however, demand for nuclear energy declined after the Three Mile Island and Chernobyl accidents, as well as from the belief that fossil energy was plentiful and would remain cheap. It took about 20 years to realize that nuclear energy would expand, owing to the energy and climate challenges the world was faced with, and with that, the need for fast reactors became obvious in order to account for the constraints of such expansion. Currently, however, the context has changed since the 1970s, and the development of fast reactors needs to be made on a new basis, taking into account new criteria linked to economy, safety, reliability, resource saving, waste minimization and physical protection against terrorism or proliferation. Such huge technological challenges also require that the new fast reactor designs be developed internationally, within multinational cooperation frameworks. Such is the goal of the Generation IV International Forum (GIF), which is a gathering of the major key actors in the field of R and D, cooperating for the sustainable development of nuclear energy. A new way of thinking has emerged from this new context: the awareness

  19. Fabrication and characterisation of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    India is pursuing its three stage nuclear power programme. In 2nd stage it plans to erect number of fast reactors which will be requiring large amount of plutonium as one of the important feed materials. It plans to begin with Mixed Oxide (MOX) fuelled fast reactors and finally shift to metallic fuel based fast reactors. Mixed oxide (MOX) (U, PU)O2, and metallic (U, Pu, Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity, low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion which is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burn up, fuel cladding interaction and lower margin between operating and melting temperature

  20. Testing stand for cosmic gas-cooling fast reactor's sample

    International Nuclear Information System (INIS)

    For carrying out of technical decision and nuclear, radiation and technological safety of gas-cooling space nuclear power plants is elaborating gas-cooling fast reactor's testing stand. In the base of its draft is taken conception of the reactor with filling up type reactor core on the base of ball fuel elements and radial coolant flowing. On the testing stand would suggested carrying out testing for study neutron and physical parameters of gas-cooling reactor, its behaviour under accident simulation. In the reactor core will suggest use carbon nitrides fuel elements with tungsten cover, provides under nominal regime relatively low fission products yield to first contour of device. Construction of fuel element was carrying out on reactor and non reactor testing and its calculated on working resource about 3000 hours. Constructive materials of reactor core have lower melting temperature, that provides organized in good time remove fuel element to containers placed under reactor in case connected with hypothetical accident. In the construction of reactor for seen tree-contours system of heat transfer and its provides multistage system of barriers against fission products yield to environment. tabs.1