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Sample records for bn-600 reactor

  1. BN-600 and BN-350 reactors

    International Nuclear Information System (INIS)

    The nuclear power plant (NPP) BN-600 has been operating since 1980 as the Beloyarsk-3 power plant. The NPP construction cost was ∼ 312 million Rubles [1980] [approximately 620 million US$ (1980)]. The planned budget was exceeded by less than 5%. First criticality was reached on 26 February 1980. The basic result of the physical startup in March 1980 (213 low (21%) enrichment fuel subassemblies (FSAs), 143 high (33%) enrichment FSAs and 13 permanent reactivity compensators) showed that the measured physical characteristics of the reactor were correspondent with the design values. Measurement of sodium flow through each FSA was carried out two times: before and after the power startup of the reactor

  2. Operating Experience with the BN-600 Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The report considers the main design features of the BN-600 liquid metal fast reactor. The performance indicators achieved for 32 years of operation are given. The measures taken to enhance BN-600 reactor power unit safety and replace and extend its equipment lifetime and their results allowed the design lifetime of the power unit to be extended up to 40 years (until 31 March 2020) are presented. The considered integrated material, methodological and theoretical investigations justifying the serviceability of the irreplaceable components of the BN-600 reactor facility have shown that the strength conditions have not been violated in any of the critical reactor components after 45 years of operation. The results, both of the actions taken to enhance the BN-600 reactor power unit safety and corrective measures related to the events at the Fukushima nuclear power plant, allow the safety of the power unit exposed to any possible extreme external impact to be improved. (author)

  3. 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

  4. Thermal stratification of sodium in the BN 600 reactor

    International Nuclear Information System (INIS)

    The signs of thermal stratification of sodium in the BN 600 reactor upper plenum revealed by the analysis of standard temperature sensors' readings are defined. The initial conditions for existence of different temperature sodium layers are given. Two approaches for realizing on a computer of equations describing sodium motion in the upper plenum of the reactor are presented. (author)

  5. 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.)

  6. Analysis of the BN-600 reactor subassemblies operating experience

    International Nuclear Information System (INIS)

    The paper presents results of the analysis of data on the deformation of the BN-600 reactor irradiated fuel subassemblies hexagonal ducts which are made of austenitic cold-worked and ferritic-martensitic steels, and of data on the deformation of fuel pin cladding which is made of austenitic cold-worked steels. Some results of subassemblies withdrawal force measurements are also presented. (author)

  7. Neutron design of BN 600 type reactor with plutonium fuel

    International Nuclear Information System (INIS)

    Briefly described is the neutron physics design of a fast reactor of the BN 600 type burning plutonium fuel. The basic specifications of the reactor are given prior to steady-state refuelling and after it. Also presented are the indices of the fuel cycle, such as the balance of heavy isotopes during refuelling and data on fuel burnup. Computations of the reactivity of one compensation fuel assembly were made for a homogenized fuel assembly in the central part and the efficiency studied of the whole system of compensation. (B.S.)

  8. Operating experience with Beloyarsk fast reactor BN600 NPP

    International Nuclear Information System (INIS)

    The main results of the seventeen-year operation of the BN600 Nuclear Power Plant are considered. The principal backfittings of the main BN600 Power Plant equipment are presented and summarised. (author)

  9. Irradiation of the experimental fuel assemblies with uranium-plutonium fuel in the BN-600 reactor

    International Nuclear Information System (INIS)

    Design features of experimental fuel assemblies (EFA) with uranium-plutonium mixed oxide fuel specific aspects of their arrangement within the BN-600 reactor core, conditions and basic results of EFA with the fuel mentioned in the BN-600 reactor are described

  10. Measurement of coolant flowrate through the fuel assemblies in BN 350 and BN 600 reactors

    International Nuclear Information System (INIS)

    Methods of the primary circuit coolant flowrate measurement in BN 350 and BN 600 reactors are described. Flowmeter design and parameters are outlined. Flowmeter application during reactor, conditions and the results of measurement are presented. Details of the modified flowmeter to be used in BN 600 reactors, that enables its verification during reactor operation by the correlation method have been briefly treated. (author). 1 ref., 1 fig

  11. Dimensional changes in elements of the BN-600 reactor core

    International Nuclear Information System (INIS)

    The spread of the data concerning dimensional changes of the components of the BN-600 core is typical for most of the construction materials but cannot be explained exclusively by their nonuniform operational parameters. The spread is caused by nonuniformity of the composition, structure, and properties of the materials within the effective specification requirements. The nonuniformity manifests itself from one lot to another and also within a single lot and, possibly, even in individual finished articles. Embrittlement of the materials during their irradiation is still another important factor which makes it impossible to obtain a burnup in excess of 10% of the heavy atoms. Proper functioning of the fuel assemblies is also limited by the stress caused by irradiation during operation as a consequence of the combined deformation of a fuel-element bundle and the hexagonal box. This holds particularly for fuel assemblies having boxes of weakly swelling ferritic-martensitic steel and fuel cladding tubes made from austenitic steel. It seems that this is the main reason for the fact that proper functioning of standard fuel assemblies with fuel cladding tubes made from cold-worked steel and boxes which is restricted to a burnup of 11-13% of the heavy atoms. When the combined deformation occurs, 05Kh12N2M is the material to be preferred for the boxes. Increasing the service life of the core components of fast reactors must involve work on improving the materials to obtain well-reproducible properties as these determine the initial state of the steel

  12. Forming the BN-600 reactor core model using GEFEST code fuel archive for SYNTES code

    International Nuclear Information System (INIS)

    The article describes the first stage of forming SYNTES code simulation model of the BN-600 reactor core, i.e. organization of transfer of the existing model of the core from GEFEST code fuel archive to a temporal database

  13. Two-dimension calculation of proposed benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This paper presents primary calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor. The analysis in this paper uses a R-Z homogeneous model of the BN-600 reactor. Calculation results include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution

  14. Three dimension calculation of proposed benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This paper presents primary calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor. The analysis in this paper uses a HEX-Z homogeneous model of the BN-600 reactor. Calculation results include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution

  15. Experience of fuel loading formation in the BN-600 reactor core

    International Nuclear Information System (INIS)

    Experience of the fuel loading formation in the BN-600 reactor core is analysed because the safety, reliability and stability of the core operation determine the power unit operation as a whole. For substantiation of the reactor operation safety every fuel loading is planned and realized by means of calculated data base. The GEFEST certified program complex is used as the permanent program for the BN-600 operation calculations

  16. Status and possibility of fuel and structural materials experimental irradiation in BN-600 reactor. Stages of BN-600 reactor core development

    International Nuclear Information System (INIS)

    The results of the irradiation of standard and experimental fuel subassemblies (SA) in BN-600 reactor are presented. The prospects of further tests on experimental SAs and on standard SAs up to 12% h.a. burnup and damage doses ≥ 90 dpa are also analyzed. (author)

  17. Main results of BN-600 reactor stress-strain state investigations

    International Nuclear Information System (INIS)

    The development of BN-600 fast reactor plant needed the solution of a series of complex engineering problems including ones for confirming integrity of the most vital structural components. The particular attention was given to the main vessel since reactor availability end safe operation of the plant as a whole depend on vessel strength end integrity. The present report deals with the main results of theoretical and experimental investigations of the stress-strain state of BN-600 reactor vessel carried out during design, start-up and initial bringing the reactor to power

  18. Bayes diagnostic system to locate the defected fuel assembly zone on BN-600 reactor

    International Nuclear Information System (INIS)

    This paper presents the method, the algorithm and the software designed to locate the zone of BN-600 reactor core containing defected fuel assembly. The BN-600 reactor is a sodium cooled fast reactor operated at Beloyarskaya NPP (Russia). The location method is based on comparison between pre-calculated and measured activity of reference radioactive nuclides in the blanket and in the primary sodium coolant. The computing algorithm is built upon the Bayesian statistical decision-making strategy under uncertainty conditions. The software environment is Dyalog APL. (authors)

  19. Power distribution control in BN-600 reactor by method of gamma-scanning of fuel assemblies

    International Nuclear Information System (INIS)

    Acceptability, convenience and reliability of γ-scanning of fuels assembles at fast reactor NPP have been analyzed and demonstrated. Error of the procedure is amount 3-6% for different fuel assemblies. The procedure is recommended as optimum one for the constructed BN-800 and perspective fast reactors. Findings allow conclusion on the accordance of BN-600 fuel assemblies powers with design parameters and insignificant (in the limits of observation accuracy) changing power distribution in new BN-600 01M2 reactor core. Experimental procedure is modernized and optimized, three cycles of measurement are realized, new experimental data on the character of radial and axial distributions of neutron field are received

  20. Operating experience of BN-600 fast neutron reactor and BN-800 reactor design

    International Nuclear Information System (INIS)

    Full text: Experience gained in Russia (USSR) in R and D work in the area of sodium cooled fast reactors in the period of 1950-1970s has been used in the design of NPP with the BN-600 reactor. Since its start-up in 1980, BN-600 reactor has demonstrated operating characteristics, which are unique for this nuclear technology. Average load factor value for 23 years of operation is near 74%, its values in 2002 and 2003 being respectively 77.35% and 75.7%. Release of inert radioactive gases is within 0.3% of reference value, while average collective dose rate of personnel is about 0.3 man. Sv per year. In the course of operation of NPP with the BN-600 reactor, effectiveness of steam generator protection system was demonstrated in 12 cases of small and large water-into-sodium leaks. Besides, unique experience was gained in confining either radioactive and non-radioactive sodium fires in case of sodium leaks from the circuits. Radioactive sodium leak from the primary auxiliary circuit occurred in December 1994 is a typical example. Total amount of sodium released from the circuit was about 1000 kg, and protection system was capable of confining sodium ignition nucleation site and limiting radioactivity release to the atmosphere by 10 Ci value. This release has had almost zero effect on radiological conditions of the NPP controlled area. BN-800 reactor design is the next stage of development of sodium cooled fast reactor technology. Fourth power unit with the BN-800 reactor is now under construction on Beloyarskaya NPP site. Innovative design approaches have been used in the BN-800 reactor in order to further improve safety of fast reactors with sodium coolant. Among these innovations are as follows: Additional 'passive' safety system using three absorber rods hydraulically suspended by the sodium flow; Passive decay heat removal system using sodium-air heat exchangers; Device for collection and retaining of the core debris in case of its disruption under conditions of

  1. Experience in the Operation of Adjustable Electric Drives of Main Circulation Pumps of BN-600 Reactors

    International Nuclear Information System (INIS)

    Adjustable electric drives of the main circulation pumps of the BN-600 reactor of the Beloyarskaya NPP have a unique layout of asynchronous valve cascades (AVC). Twenty five years of successful operation of such drives prove the expediency of the use of AVC, which deserves study and application in the design of power units with BN-800 reactors

  2. Development and introduction of automated lines for fabrication of vibrocompacted fuel elements for BN-600 reactor

    International Nuclear Information System (INIS)

    In the framework of international program modernization of technological complex for fabrication granular, suitable for vibration compacting fuel, fuel elements and fuel assemblies is realized. The aim of modernization is to provide BN-600 reactor with MOX fuel on the basis of weapon plutonium

  3. Plutonium disposition in the BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant

    Science.gov (United States)

    Moses, D. L.; Chebeskov, A. N.; Matveev, V. I.; Vasiliev, B. A.; Maltsev, V. V.

    In 1996, the United States and the Russian Federation completed an initial joint study that evaluated the candidate options for the disposition of surplus weapons-derived plutonium in both countries. While Russia advocates building new reactors for converting weapons-derived plutonium to spent fuel, the cost is high, and the continuing joint study of the Russian options is considering only the use of the existing VVER-1000 LWRs in Russia (and possibly in Ukraine) and the existing BN-600 fast-neutron reactor at the Beloyarsk Nuclear Power Plant in Russia. The BN-600 reactor, which currently uses enriched uranium fuel, is capable with certain design modifications of converting up to 1.3 metric tons (MT) of surplus weapons-derived plutonium to spent fuel each year. The steps needed to convert BN-600 to a plutonium-burner core will be discussed. The step involving the hybrid core allows an early and timely start that takes advantage of the limited capacity for fabricating uranium-plutonium mixed-oxide fuel early in the disposition program. The design lifetime of BN-600 must safely and reliably be extended by 10 yr to at least 2020 so that a sufficient amount of plutonium (˜20 MT) can be converted to spent fuel.

  4. Plutonium disposition in the BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant

    International Nuclear Information System (INIS)

    In 1996, the United States and the Russian Federation completed an initial joint study that evaluated the candidate options for the disposition of surplus weapons-derived plutonium in both countries. While Russia advocates building new reactors for converting weapons-derived plutonium to spent fuel, the cost is high, and the continuing joint study of the Russian options is considering only the use of the existing VVER-1000 LWRs in Russia (and possibly in Ukraine) and the existing BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant in Russia. The BN-600 reactor, which currently uses enriched uranium fuel, is capable with certain design modifications of converting up to 1.3 metric tons (MT) of surplus weapons-derived plutonium to spent fuel each year. The steps needed to convert BN-600 to a plutonium-burner core will be discussed. The step involving the hybrid core allows an early and timely start that takes advantage of the limited capacity for fabricating uranium-plutonium mixed-oxide fuel early in the disposition program. The design lifetime of BN-600 must safely and reliably be extended by 10 yr to at least 2020 so that a sufficient amount of plutonium (∝20 MT) can be converted to spent fuel. (orig.)

  5. Release of radioactive fission products from BN-600 reactor untight fuel elements

    International Nuclear Information System (INIS)

    The experimental data on the release of radioactive fission products from BN-600 reactor untight fuel elements are given in the report. Various groups of radionuclides: inert gases Xe, Kr, volatile Cs, J, non-volatile Nb, and La are considered. The results of calculation-experimental study of transfer and distribution of radionuclides in the reactor primary circuit, gas system and sodium coolant are considered. It is shown that some complex radioactivity transfer processes can be described by simple mathematical models. (author)

  6. Service properties of structural materials of BN-600 reactor fuel assemblies at high damaging doses

    International Nuclear Information System (INIS)

    Based on postradiation investigation of fuel assembly materials of BN-600 reactor a consideration is given to main ways of designing materials which can provide for high burn-up of nuclear fuels in fast reactors. Austenitic steels 08KhN11M3T, 10Kh17N13M2T and ferrite-martensitic steels 1Kh13M213FR, 05Kh12N2M were tested as fuel assembly cans in BN-600 reactor. Austenitic steels EhJ-847, EhP-172, ChS-68 were used for fuel cans. It'is shown that radiation resistance of the steels can be improved by optimization of chemical composition and by enhanced homogeneity of composition, structure and initial mechanical properties. 20 refs.; 4 figs

  7. European contribution to Phase 3 of the benchmark core analysis for the BN-600 hybrid reactor

    International Nuclear Information System (INIS)

    This European participation in Phase 3 of the benchmark (BN-600) analysis consist of a joint contribution from France and the UK. Calculations were performed by ERANOS code and data system which has been developed in the framework of European cooperation on fast reactors. Results are presented for all the core neutronic parameters, both for homogeneous and heterogeneous core models and both for beginning and end of fuel cycle

  8. BN-600 Reactor Capability for the Development of Fuel Pin and FSA Materials for LMFRS

    International Nuclear Information System (INIS)

    The BN-600 is the most powerful power fast reactor in the world. Uranium oxide fuel of three enrichments on U-235 is used. The FA lifetime is 560 effective full power days (EFPDs), two reloadings per year with average duration between reloadings, 140 EFPDs. The maximum damage dose rate is 41 dpa per year. Sodium temperature range in the core is 368-550oC. The present unique combination of irradiation conditions is extremely attractive to support the irradiation of materials and items for the testing purposes. The BN-600 is the power reactor with assigned commercial parameters, and the tests should not essentially influence its operation mode. Besides it is necessary to take into account normative safety rules and limits on allowed perturbations in the reactivity margin and the heat release distribution. At preparation of tests licensing the nuclear and radiation safety justification should be supported with theoretical and experimental results. The paper describes the BN-600 irradiation capabilities, irradiation test experience and requirements for the irradiation tests arrangement. (author)

  9. Increasing the economic efficiency of nuclear fuel usage in the BN-600 reactor of the Beloyarsk NPP

    International Nuclear Information System (INIS)

    Improvement of technical and economic indices of the BN-600 reactor is largely dictated by increase in the efficiency of nuclear fuel use. In the period from 1980 to 2003 two modernizations (01M and 01M1 cores) were carried out. Conversion to the 01M2 core with a four-time fuel assemblies reloading and fuel assembly life of 560 effective days was started in 2004. Major neutronic characteristics of the BN-600 reactor cores are provided

  10. Isotopic Transmutation and Fuel Burnup in BN-600 Hybrid Fast Reactor Core

    International Nuclear Information System (INIS)

    BN-600 fast reactor core was modeled using MCNPX computer code. The core configuration and material composition, for the hybrid design, was simulated in this model. The power generated in different zones was determined and the results were compared with published results and found acceptable. Isotopes transmutations in various zones were estimated. The uranium isotopes are major contributors to power production in this reactor, the probability of plutonium incineration will increase with the increase in the use of MOX oxide. The transmutation of minor actinides is not obvious in this configuration

  11. Weapons-grade plutonium effective utilization in BN-600 fast neutron reactor with vibropacked MOX-fuel

    International Nuclear Information System (INIS)

    For the last few years, within the framework on disposition of plutonium declared as not required any longer for defense programs, Rosatom organizations (Russia) have been carrying out research and design activities on utilization of this plutonium in BN-600 fast neutron reactor being operated at Beloyarskaya NPP. These activities were started with scientific-technical and financial support from USA and France. At present considerable progress in these studies is achieved with scientific-technical and financial support from Japan Nuclear Fuel Cycle Research Institute (JNC). Scenario to use the BN-600 reactor for ex-weapons plutonium utilization comprises two stages. The first stage includes the BN-600 reactor for ex-weapons plutonium utilization comprises two stages. The first stage includes the BN-600 reactor transition to hybrid core, the second stage - to the full MOX core. It is shown that using BN-600 fast neutron reactor for ex-weapons plutonium utilization is the real way in Russian situation to convert weapons material into material that can not be used again in nuclear weapons. (author)

  12. Joint European contribution to phase 4 of the BN-600 hybrid reactor benchmark core analysis

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fuelled core of BN-600 reactor was endorsed as an international benchmark. Phase 4 of the RCM benchmark studies consider full MOX core differentiated by design measures to reduce the sodium void worth. Parameters to be calculated were: fuel and steel Doppler coefficients; fuel density coefficient; sodium density coefficient; power distribution for fuel and non-fuelled regions; β effective and prompt neutron life time. Heterogeneity effects are evaluated. Analysis was carried out using ERANOS code and data system for fast reactors. Nuclear data library is based on JEF2.2. Accurate calculations of control rod heterogeneity effects with homogeneous equivalent cross sections for control rod absorbers were prepared using reactivity equivalence technique

  13. Trial use of JOKER software package for justification of safe the BN-600 reactor operation under transients

    International Nuclear Information System (INIS)

    The article presents Beloyarsk NPP work on the development of JOKER software package database of actual equipment of the BN-600 reactor designated for justification of the reactor safe operation under transients. An example of calculation of actual parameters of primary sodium pump and a fragment of equipment database are given

  14. Thermohydraulic characteristics of the BN-600 reactor at the Beloyarsk nuclear power station

    International Nuclear Information System (INIS)

    During the period of startup and adjustment of the BN-600 reactor, when its rated power was reached, the hydraulic characteristic of the primary circuit was determined by measuring the flow rate of the sodium coolant through the fuel assemblies and the control-rod assemblies. A flow-measuring device was used, consisting of a built-in magnetic flowmeter with a calibration characteristic obtained on a sodium test stand. Processing of the measurement data indicated that the hydraulic characteristic of the primary circuit and the distribution of coolant flow rate over the throttling zones are in good agreement with the design data. Repeatability of the results is observed for flow-rate measurements through the fuel assemblies in the same cells during different microperiods of operation

  15. Methodology and results of operational calculations of fuel temperature in fuel elements of the BN-600 reactor fuel assemblies

    International Nuclear Information System (INIS)

    The article presents methodology of peak fuel temperature determination and computational investigations of fuel temperature condition in fuel elements of fuel assemblies of various types during the BN-600 reactor operation. The effect of sodium uranate in the gap between fuel and cladding of the fuel element on the heat transfer processes is considered

  16. On the possibility of sodium boiling detection in the BN-600 reactor by neutron noises

    International Nuclear Information System (INIS)

    The possibility of early diagnostics of sodium leakage and its boiling in the BN-600 reactor fuel assembley on the basis of neutron flux or reactivity noise analysis is studied. So determine the nature of integral and local neutron flux changes under fuel assembley blockage calculated and experimental data, obtained at the BOR-60 reactor, are analysed. Calculations are carried out using the NF-6 program complex. The reaction of local neutron flux monitors, made of rhodium, was determined during the experiments besides reactivity change Δk/k measuring. It is ascertained, that the effect of a fuel assembly complete devastation depending ion its location in the zone changes within the range from -10 Δk/k up to 2x10-5 Δk/k. The amplitude of signal pulsations of the neutron flux monitor, installed on the turning plus bottom plate, is 0.25-0,51%, taking into account its dynamic characteristics. It means, that using one of such monitors it is possible point of the reactor core. So register a weaker boiling a number local monitors will be needed

  17. Joint European contribution to phases 1 and 2 of the BN600 hybrid reactor benchmark core analysis

    International Nuclear Information System (INIS)

    This paper describes the ERANOS code developed within the European cooperation on fast reactors. Reference scheme and ERANOS code validation are included. The method for BN-600 reactor core analysis and the results of phases 1 and two are presented. They include effective multiplication factors, fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution; beta and neutron life time; reaction rate distribution

  18. Progress of demonstration experiment on irradiation of vibro-packed MOX fuel assemblies in the BN-600 reactor

    International Nuclear Information System (INIS)

    According to the Concept of Russian Federation MINATOM and DOE of USA, plutonium to be released as a result of nuclear weapon dismantling is to be used in nuclear power engineering - in the form of MOX fuel in fast or thermal reactors. The scenarios of Russian weapon-grade plutonium disposal provide for its application as a MOX fuel in the hybrid core of BN-600 (BNPP) and in the BN-800 reactor under construction. The following procedures developed at JSC 'SSC RIAR' can be taken as basic ones: - pyroelectrochemical granulation of uranium-plutonium oxides resulting in granulated MOX fuel production; - vibropacking of granulated fuel directly in the fuel pin cladding. Experience in vibro-packed fuel tests in the BOR-60, BN-350 and BN-600 reactors showed that vibro-packed MOX fuel had acceptable service life parameters even at super high burnup (about 30% h.a.). The demonstration experiment has been conducted since 2004 within the framework of international cooperation between Russian and Japanese organizations: RIAR, IPPE, OKBM, BNPP, MEXT, JAEA, and PESCO. The goal of the experiment was to validate the possibility of vibro-packed MOX fuel assembly (FA) application for weapon-grade plutonium disposal in a fast reactor Under the program of the demonstration experiment, RIAR conducted pyroelectrochemical plutonium conversion into granulated MOX-fuel. The granulated fuel was used for manufacturing fuel pins by vibropacking and the fuel pins were assembled into 21 experimental fuel assemblies (EFAs) to be tested in BN-600. OKBM developed the EFA design. Besides, OKBM prepared the detailed program of tests and substantiated irradiation parameters in cooperation with IPPE, and also provided licensing of irradiation in the BN-600 reactor in cooperation with BNPP. BNPP provided irradiation of EFAs. All EFAs were irradiated during the specified life in accordance with the program of tests. The maximum liner heat rate of fuel pins in the FA were in the range of 39.5-45.3 k

  19. State of fuel elements in BN-600 reactor fuel assembly 917137489 when reaching maximal damaging dose of 93.7 dpa

    International Nuclear Information System (INIS)

    A complex of post-irradiated studies was accomplished for BN-600 reactor fuel cans of steel ChS-68. The studies were carried out by methods of profilometry, metallography short-term mechanical tests after irradiation up to damaging dose of 93.7 dpa. High volumetric changes, zero plasticity, corrosion defects testify to the fact that further operation of the steel under conditions of BN-600 reactor up to damaging doses more than 91-93 dpa is inadmissible

  20. Progress of Demonstration Experiment on Irradiation of Vibro-packed MOX Fuel Assemblies in the BN-600 Reactor

    International Nuclear Information System (INIS)

    The paper presents progress results, including fabrication of vibro-packed MOX fuel pins and 21 fuel assmblies (FAs) for fast reactor BN-600, irradiation parameters and postirradiation examination (PIE) results. It is shown that no violations of safe operation limits took place. The activities within the framework of the demonstration experiment are based on the international cooperation and have been performed with the support and participation Russian and Japanese organizations; RIAR, IPPE, OKBM, BNPP, JAEA and PESCO. The goal of the experiment is to validate possibility of using vibro-packed MOX FA for weapon plutonium disposition in the fast reactors. (author)

  1. Results of studies on safety of the BN-600 reactor with hybrid core for the purpose of weapons Pu disposition

    International Nuclear Information System (INIS)

    BN-600 fast neutron reactor is considered as a part of the Russian program on weapons plutonium utilization. For this purpose it is proposed to use hybrid core with partial loading of MOX fuel based on weapons grade plutonium. In this view, safety analysis of the BN-600 reactor with the hybrid core has become necessary. This analysis is carried out in accordance with the Russian regulatory documents on the NPP safety. Initial studies on some issues in this area were made in cooperation with France in 1995. These studies were continued to larger scale in 1997 with support from the USA. The most significant work was carried out during 2001-2004 period with support from and active participation of Japan, and its results are presented in this paper. Analytical studies included preparation of input data, analysis of abnormal operating conditions of the reactor, design basis and beyond design basis accidents and probabilistic safety analysis of the power unit. Results of analysis confirmed sufficient effectiveness of safety systems as applied to modified reactor design and observation of regulatory requirements in case of beyond design accident. In the course of safety analysis, no problems caused by vipac MOX fuel were encountered. (author)

  2. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  3. 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)

  4. Prospects for improvement of supporting systems of BN reactors based on BN-600 and BN-800 engineering experience

    International Nuclear Information System (INIS)

    Full text: Spent fuel assembly ablution system is one of system which are only used for BN reactors. Besides sodium ablution in this system, a Spent fuel assembly is tested for leakiness with the help of the defective fuel assembly detection system (DFADS-AC). Summary of system work regime (process). Ablution cell (AC) is filled with nitrogen. Spent fuel assembly is placed in SA. 1. Ablution regime in case when there are no suspicions of faulty sealing based on other shell hermeticity control readout. Steam ablution is conducted. After that SA is filled with nitrogen again. Spent fuel assembly is kept in nitrogen for heating-up due to residual heat. Then nitrogen is delivered from SA to gas DFADS-AC. If no fault is found in the sealing, spent fuel assembly conducted is washed by water which also delivered for control to water DFADS-AC. 2.Ablution regime in case when there are suspicions of faulty sealing based on other cover hermeticity control readout. First of all the control is performed in gas DFADS-AC. If there is no faulty sealing indication the series of ablution scheme is the same as described above. The main feature of spent fuel assembly ablution system of BN-600 is that one DFADS-AC is used for two ACs. It means that the parallel work of two ACs is impossible. Besides pipelines for delivery of steam and nitrogen to AC are united, which result is steaming of nitrogen pipeline. This leads to extra difficulties in parallel usage of two ACs. Changes in the spent fuel assembly ablution system for BN-800 based on operation experience of BN-600: There are two gas DFADS-AC and water DFADS-AC, one for each AC; Separation of pipelines for delivering of steam and nitrogen to SA; Possibility of separate delivery of ablution substance to each SA; There is no usage of gas DFADS-AC in nitrogen drop to the technology blowing system. All performed changes in the current system make it possible to perform spent fuel assembly ablution in two AC simultaneously. It means

  5. Main regularities in variations of mechanical properties and microstructure of fuel element assembly can material (steel EhP-450) irradiated in BN-600 and BN-350 reactors

    International Nuclear Information System (INIS)

    The complex of mechanical properties of steel EhP-450 fuel assembly cans irradiated in fast reactors was under study. The steel is shown to possess a high resistance to swelling as well as acceptable values of mechanical properties under tension and impact bending. Based on the results obtained a conclusion is made that in a low-temperature zone of BN-600 reactor fuel assembly cans at 15% burnup the most essential change in mechanical properties should be expected in the vicinity of a lower reactor core boundary at damaging doses of 20-40 dpa

  6. Investigation of the performance of uranium-plutonium mixed oxide fuel pins for the reactor BN-600

    International Nuclear Information System (INIS)

    A subassembly of experimental fuel pins of the type to be used in BN-600 and using uranium-plutonium mixed oxide fuel had been irradiated up to burnup of 11% heavy atoms and an integrated neutron dose of 6.6 x 1022 n/cm2 (E>=0,1 MeV) at a maximum clad temperature of 6800C. The most important results were as follows: 1) The dimensional change of the hexagonal wrapper tube (steel OCh18N10T) can be attributed to irradiation induced steel swelling of 7,6 vol.% at the neutron dose given above. 2) On an average the diameter of the fuel pins increased by 2% with a maximum of 3%. This diameter increase is essentially due to steel swelling. The diametral cladding deformation caused by radiation creep did not exceed 0,3%. 3) The maximum swelling of the clad material (solution treated steel OCh16N15M3B) was found to be 7,3% at about 5000C and at a neutron dose of 6.6 x 1022 n/cm2 (E>=0,1MeV). 4) The corrosive interaction between the mixed oxide fuel, the initial O/M-ratio of which was 1.97 to 1.98, and the steel clad was negligible and had no influence on the performance capability of the fuel clad. 5) Strength and plasticity of the clad were reduced due to the irradiation, especially in the high temperature region, but the clad kept its function. 6) The fraction of released fission gas is independent from the initial fuel density and was here found to be approx. 80%. 7) The fuel pins of the BN-600 type with U-Pu-mixed oxide (0.85 UO2 - 0.15 PuO2) proved to perform satisfactorily up to a burnup of 11%. (orig.)

  7. BN-600 fuel elements and fuel assemblies operating experience

    International Nuclear Information System (INIS)

    Consideration is given to the data on fuel burnup of standard fuel assemblies of the BN-600 reactor first core charge and that for modified core; data on operation ability of fuel assemblies of the first charge type are given. Data on main results of primary post-irradiation examination of fuel assemblies and fuel elements and maximal values of fuel burnup, achieved in particular fuel assemblies of BN-600 reactor are presented. 4 figs.; 1 tab

  8. Use of different programs for calculating the flux density of neutrons activating sodium in the secondary circuit of a NPP with the BN-600 reactor

    International Nuclear Information System (INIS)

    Possibilities of application of the RADAR, TVK-2D and MMKFK program complexes to calculate the BN-600 type reactor shields are analyzed. TVK-2D program (ALGOL-DDR, BESM-6 computer) is designed for two-dimensional calculations of reactors in diffusion multigroup finite-difference approximation using classical and unified perturbation theory. The RADAR system (FORTRAN-4, BESM-6 computer) realizes Boltzmann equation solution by iterative synthesis method in multigroup diffusion approximation. The MMKFK complex (FORTRAN, BESM-6 computer) is used to calculate radiation transport in reactors and cells. The complex is improved: at large ratioes of neutron flux attenuation the methods of splitting and roulette are realized. Calculational results of the integral by energy and mean by zones values of neutron flux density in radial shield and sodium activity in the secondary coolant circuits are presented. Good conformity of the data obtained is pointed out. Conclusion is made about the applicability of the program systems investigated to calculate fast reactor shields at different stages of design. The RADAR system due to its quick operation will be more efficient at the initial stages, while the MMKFK system - at final ones, when high accuracy of calculation is required

  9. Performance of the BN-600 reactor fuel pins with claddings made of austenitic steels EI-847, EP-172 and ChS-68 at high radiation damage levels

    International Nuclear Information System (INIS)

    Development of austenitic stainless steels for fuel pin cladding in fast reactors, capable to provide their reliable and economic operation, is one of the most important problems of reactor materials science. Intensive works aimed to increase the fuel burn-up in the BN-600 reactor were conducted for many years under the leadership of VNIINM. These works included the development of new cladding steels. A set of experimental subassemblies with fuel pin cladding fabricated from the new steels has been produced and irradiated in the BN-600 reactor. Characteristics of the subassemblies and main irradiation conditions are shown. Post-irradiation examinations of a part of fuel pins from these subassemblies(7-9 fuel pins) have been conducted in the IPPE hot laboratory As a result of post-irradiation examination of fuel pins which reached the maximum burn-up of 11.6% h.a., it was established, that the largest degradation of operational properties of fuel pin claddings is observed in the region of the maximum diameter increase and is reveled as a total embrittlement of the cladding material and the appearance of cracks of a substantial depth on the cladding internal surface. Processes resulting in the deterioration of fuel-pin cladding properties (embrittlement, formation of microcracks) are directly connected with swelling and/or with the radiation-induced segregation which occurs in the swelling temperature range and due to action of forces causing the swelling. The interrelation of processes of corrosion cracking and swelling of fuel pin cladding is considered. The effect of the stresses arising due to gradient of swelling in the cladding wall appears as most important. The level of stresses is also determined by the temperature dependence of swelling of steels used for fabricating the fuel pin cladding. After high dose irradiation there is a rather high level of residual stresses in fuel pin cladding that leads to failure of fuel pins during manipulations with them in

  10. Fuel release into primary sodium of the BN-600 and BR-10 reactors through natural and artificial defects in fuel pin claddings

    International Nuclear Information System (INIS)

    The most interesting on this problem stage from 1981 to 1987 of the BN-600 reactor operation with failed fuel pin claddings is described. The assessment is presented of total number of nonhermetic fuel pins in the core during this period, having contact between fuel and coolant. Approximate number of these fuel pins is assessed to be 60. According to estimates, the average fuel release from damaged fuel pins into sodium with contact between fuel and coolant, was 0.2-0.5%. Similar assessments of Nb-95 activity in deposits on the primary equipment surfaces show, that relative release of solid fission products from damaged fuel pins, having contact between fuel and coolant, is of the same order with relative fuel loss value. In this paper the results are presented of investigations at the BR-10 reactor of fuel release from fuel pins with natural and artificial defects in claddings. The fuel release from fuel pins with natural defects was considered for the core standard SAs with plutonium dioxide. It has been found, that from a breathed fuel pin, having contact of fuel with coolant, during a cycle up to 0.25% of fuel released to sodium. The results are discussed of a large series of experiments at the BR-10 on fuel pin behaviour, in claddings of which through defects were made. (author)

  11. Post-irradiation examination of Ti or Nb stabilized austenitic steels irradiated as BN-600 reactor fuel pin claddings up to 87 dpa

    International Nuclear Information System (INIS)

    The results of postirradiation study of fuel pins with claddings fabricated from the 16Cr-15Ni-3Mo-Nb (EI-847), 16Cr-15Ni-3Mo-Nb-B (EP-172) and 16Cr-15Ni-2Mo-Ti-V-B (ChS-68) austenitic stainless steels in 20% cold-work condition are given. All fuel pins after irradiation in the BN-600 reactor to peak burn up of 11.6% (displacement dose of 83 dpa) and remained its tightness. At the same time, a number of fuel pins have failed during low-load handling in hot cells. Tensile mechanical tests revealed a drastic decrease in strength and a severe embrittlement of the cladding material taken from some parts of fuel pins. For these parts numerous deep microcracks at the inner surface of pin cladding have been observed. Locations of the maximum cladding property degradation coincides with locations of the peak diameter increase and peak swelling. The effects of high swelling and radiation-induced segregation on mechanical properties and corrosion resistance of the fuel pin cladding are discussed. (author)

  12. Licensing support experience of the BN-600 operation

    International Nuclear Information System (INIS)

    The experience gained by the Russian regulatory body for licensing support of the operation of sodium-cooled fast reactor (Beloyarskaya nuclear power plant BN-600) from the standpoint of further evolution of the sodium-cooled fast reactors are described. For more than thirty summers period of the commercial operation of the BN-600 the regulatory body has fulfilled safety reviews of the wide range of justification report concerning of technical decision of the implemented modernizations which were carried out on the power unit to increase the technical and economic reactor indicators. Accident-free operation of the BN-600 reactor evidences both the quality of the development and level of mastery of this reactor technology, and performance of the appropriate supervision from the regulatory body. (author)

  13. BN-600 full MOX core benchmark analysis

    International Nuclear Information System (INIS)

    As a follow-up of the BN-600 hybrid core benchmark, a full MOX core benchmark was performed within the framework of the IAEA co-ordinated research project. Discrepancies between the values of main reactivity coefficients obtained by the participants for the BN-600 full MOX core benchmark appear to be larger than those in the previous hybrid core benchmarks on traditional core configurations. This arises due to uncertainties in the proper modelling of the axial sodium plenum above the core. It was recognized that the sodium density coefficient strongly depends on the core model configuration of interest (hybrid core vs. fully MOX fuelled core with sodium plenum above the core) in conjunction with the calculation method (diffusion vs. transport theory). The effects of the discrepancies revealed between the participants results on the ULOF and UTOP transient behaviours of the BN-600 full MOX core were investigated in simplified transient analyses. Generally the diffusion approximation predicts more benign consequences for the ULOF accident but more hazardous ones for the UTOP accident when compared with the transport theory results. The heterogeneity effect does not have any significant effect on the simulation of the transient. The comparison of the transient analyses results concluded that the fuel Doppler coefficient and the sodium density coefficient are the two most important coefficients in understanding the ULOF transient behaviour. In particular, the uncertainty in evaluating the sodium density coefficient distribution has the largest impact on the description of reactor dynamics. This is because the maximum sodium temperature rise takes place at the top of the core and in the sodium plenum.

  14. GIDROPRESS development center experience in the field of extension of the BN-600 NPP steam generator service life and steam generator development tasks for new fast reactor power units

    International Nuclear Information System (INIS)

    The article presents the work of the GIDROPRESS development center both on evaluation of residual life and lifetime extension of the BN-600 power unit PGN-200M steam generators. The strength of the steel constructions was justified; the recommendations were given on seismic resistance improvement. The imperfection of the standing regulatory base was noted, the measures to correct this situation were highlighted. The BN-600 steam generators positive operating experience as well as the examination results of the critical components condition after the steam generators long performance enables to confidently change over to the new generation of large steam generators, thus considerably improving technical and economic characteristics of prospective designs of fast reactors

  15. BN-600 hybrid core benchmark analyses

    International Nuclear Information System (INIS)

    Benchmark analyses for the hybrid BN-600 reactor that contains three uranium enrichment zones and one plutonium zone in the core, have been performed within the frame of an IAEA sponsored Coordinated Research Project. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN-600 core were evaluated. The comparison of the diffusion and transport results obtained for the homogeneous representation generally shows good agreement for most parameters between the RZ and HEX-Z models. The burnup effect and the heterogeneity effect on most reactivity parameters also show good agreement for the HEX-Z diffusion and transport theory results. A large difference noticed for the sodium and steel density coefficients is mainly due to differences in the spatial coefficient predictions for non fuelled regions. The burnup reactivity loss was evaluated to be 0.025 (4.3 $) within ∼ 5.0% standard deviation. The heterogeneity effect on most reactivity coefficients was estimated to be small. The heterogeneity treatment reduced the control rod worth by 2.3%. The heterogeneity effect on the k-eff and control rod worth appeared to differ strongly depending on the heterogeneity treatment method. A substantial spread noticed for several reactivity coefficients did not give a significant impact on the transient behavior prediction. This result is attributable to compensating effects between several reactivity effects and the specific design of the partially MOX fuelled hybrid core. (author)

  16. Study of the influence of pH value and passivation in nitrate solutions on corrosion resistance of reactor BN-600 cladding tubes of steel EhP-450

    International Nuclear Information System (INIS)

    A study is made into the influence of steel EhP-450 cladding tube operational conditions in BN-600 reactor, a pH value of fuel storage pool water medium and the passivation in iron and chromium nitrate aqueous solutions on corrosion resistance of the steel. The corrosion resistance is judged from the mass loss. It is established that at inner surfaces of cladding tubes the chromium-depleted and carbon-enriched layers of reduced corrosion resistance are formed in operation. Analytical expressions are obtained which allow to have quantitative correlations between corrosion characteristics (corrosion rate, corrosion products entrainment, surface layer thickness) and operational conditions for steel EhP-450 in BN-600 reactor. The studies provide support for the view that the passivation treatment of EhP-450 cladding tubes in a Fe(NO3)3 solution with the aim of their corrosion protection when holding in the water at pH ≥ 8.0 in a fuel storage pool has considerable promise

  17. Accoustic background of BN-600 steam generator

    International Nuclear Information System (INIS)

    In this paper the results of accoustic background for BN-600 steam generator in nominal operating conditions are presented. The 1-200 kHz accoustic background of evaporator and reheater modules are given

  18. Conception of the BN-600 power plant safety estimation

    International Nuclear Information System (INIS)

    A concept of safety evaluation of the Beloyarsk NPP power unit equipped with the BN-600 reactor is considered. It is suggested that both current state of the unit is evaluated control of the physical barriers integrity, control of safe operation limits and conditions) and prediction is made of possible changes in the power unite state, bearing in mind current values of process parameters and equipment operation conditions

  19. Variations of mechanical properties in steel ChS-68 cold def. under irradiation as a fuel can material in BN-600 reactor up to damaging doses of 10-40 dpa

    International Nuclear Information System (INIS)

    Mechanical properties of fuel can material (steel ChS-68, 20% cold.def.) are investigated after irradiation in BN-600 reactor for 160 and 264 EFDP up to burnups of 2.5 and 4.5% h.a. and damaging doses of 20 and 37 dpa. Tensile tests of annular specimens are carries out in the temperature range of 20-800 deg C. The results obtained show that along with a conventional decrease of plasticity due to irradiation defect formation the high temperature embrittlement is observed. The effect is the greater the higher is the neutron fluence. The effect manifests itself in the material irradiated in the temperature range of 500-600 deg C. The phenomenon of high temperature embrittlement is associated with alloying elements segregation and precipitation of noncoherent particles resulting in grain boundaries weakening

  20. Characteristics of radiation porosity formed upon irradiation in a BN-600 reactor in the fuel-element cans of cold-deformed steel EK-164 (06Kh16N20M2G2BTFR)-ID c.d.

    Science.gov (United States)

    Portnykh, I. A.; Kozlov, A. V.; Panchenko, V. L.; Mitrofanova, N. M.

    2012-05-01

    At present, it is the austenitic cold-deformed steel EK164 (06Kh16N20M2G2BTFR)-ID that is considered as a promising material for the achievement of a maximum damage (no less than 110 dpa) and maximum burnup (≥15%). In this work, we have determined the characteristics of porosity formed upon irradiation in a BN-600 reactor to the maximum damaging dose of 77 dpa in the materials of fuel-element cans made of cold-deformed steel EK164-ID c.d. A comparison has been made with analogous characteristics obtained earlier using the standard material, i.e., the cold-deformed steel ChS68 (06Kh16N 15M2G2TFR)-ID c.d.

  1. Experimental and calculating substantiation of reactivity balance and energy-release distribution in BN-600 core

    International Nuclear Information System (INIS)

    This paper summarizes the results of a series of experimental and theoretical studies carried out in 2003-2010 on the substantiation of neutron-physical characteristics of BN-600 core. This work caused BN-600 transition to the new core 01M2 with high burnup 11.2% h.a., need for analysis and comprehension of the BN-600 experience in anticipation of the end of BN-600 design life cycle and extending it to 10-15 years, development and introduction of new methods of analysis, in particular, the beginning of use for BN-600 core calculation precision method of Monte Carlo, free from most of the limitations of traditional diffusion codes. In the experiments was a change of equipment and measurement techniques. By the beginning of this work cycle energy-release of BN-600 had not been measured for about 10 years. As a result of experimental and calculating studies was created the coordinated and consistent database of neutron-physical characteristics for the fast sodium reactor of medium power. With the account of long-term operating experience of the BN-600 were finally brought and tested measurement methods of reactivity characteristics based on the new reactimeter methodology and gamma scanning of fuel assemblies. The development and application of new programs for neutronic calculations (new version of GEFEST and JARFR codes based on ABBN-93 constant system, the ModExSys system and MMKKENO code based on the Monte Carlo method) has allowed at a new level to understand and evaluate the methodological features of calculation of basic parameters of the BN-600 core. Substantiated values of the basic neutron-physical characteristics of the BN-600 core with high level of accuracy. (author)

  2. BN-600 hybrid core benchmark Phase III results

    International Nuclear Information System (INIS)

    The main objective of the CRP on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects, is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at using weapons-grade plutonium for energy production in fast reactors. BN-600 hybrid reactor taken as benchmark. Earlier, two dimensional and three dimensional diffusion theory BN-600 benchmark calculations were done. This report describes the results of the burnup and heterogeneous calculations done for the proposed BN-600 hybrid core model as a part of Phase III benchmark. BN-600 benchmark has been analyzed at beginning of cycle (BOC) with XSET98 data set and 2-D and 3-D diffusion codes. The 2-D results are compared with the earlier results using the older CV2M data set. The core has been burnt for one cycle using 3-D burnup code FARCOBAB. The burnt core parameter has also been analyzed in 3-D. Heterogeneity effects on reactivity have been computed at BOC. Relative to the use of CV2M data, use of XSET98 data results in increased magnitudes of fuel Doppler worth and sodium density worth. Compared to 2-D results , in the 3-D results, the Keff is lower by about 220 pcm, sodium density worth is higher by about 30% and steel density worth becomes nearly zero or small positive from a negative value in 2-D. The conversion ratio at BOC is 0.669 as computed in 3-D. The burnup reactivity loss due to 140 days at full power (1470 MWt) is 0.0252. The conversion ratio at end of cycle (EOC) is 0.701. The other parameters have been estimated with SHR up condition as desired in the phase III benchmark specifications. Fuel Doppler worth is 7% more negative, sodium density worth is 16% less positive and steel density worth is more negative at EOC compared to BOC. Absorber rod (SHR) worth is higher by 4.9 % at EOC. Heterogeneity effect (core and SHR combined) on multiplication factor is small. For mid SHR

  3. BN-600 full MOX core benchmark analysis (PHYSOR 2004 paper)

    International Nuclear Information System (INIS)

    As a follow-up of the BN-600 hybrid core benchmark, a full MOX core benchmark was performed within the framework of the IAEA co-ordinated research project Discrepancies between the values of main reactivity coefficients obtained by the participants for the BN-600 full MOX core benchmark appear to be larger than those in the previous hybrid core benchmarks on traditional core configurations. This arises due to uncertainties in the proper modelling of the axial sodium plenum above the core. It was recognized that the sodium density coefficient strongly depends on the core model configuration of interest (hybrid core vs. fully MOX fuelled core with sodium plenum above the core) in conjunction with the calculation method (diffusion vs. transport theory). The effects of the discrepancies revealed between the participants' results on the ULOF and UTOP transient behaviours of the BN-600 full MOX core were investigated in simplified transient analyses. Generally the diffusion approximation predicts more benign consequences for the ULOF accident but more hazardous ones for the UTOP accident when compared with the transport theory results. The heterogeneity effect does not have any significant effect on the simulation of the transient The comparison of the transient analyses results concluded that the fuel Doppler coefficient and the sodium density coefficient are the two most important coefficients in understanding the ULOF transient behaviour. In particular, the uncertainty in evaluating the sodium density coefficient distribution has the largest impact on the description of reactor dynamics. This is because the maximum sodium temperature rise takes place at the top of the core and in the sodium plenum

  4. Calculation of the BN-600 fuel assemblies mode in a gas medium

    International Nuclear Information System (INIS)

    Potentiality of calculated modeling of temperature conditions of warming up elements of spent fuel assemblies of the BN-600 reactor during their transportation within gaseous medium is shown. The calculated modeling of spent fuel assemblies warming up in gaseous medium, their residual heat release values being different, permits substantiating and optimizing safe conditions of post-reactor handling of the fuel assemblies

  5. Licensing Support Experience of the BN-600 Operation

    International Nuclear Information System (INIS)

    License procedure - Main principle: • All works, including fatigue tests of new types of fuel, are carried out at the unit 3 Beloyarsk nuclear power plants with the BN-600 reactor with the justification of the regulatory body. • Justification procedure is standard for all power units and independent from the reactor types. • The regulatory body and independent experts or technical support organizations, which can be involved in this work by the regulatory body, review SAR, operational manuals and other operator documents. • Safety requirements (i.e. Federal rules and codes). The project and design documents shall meet safety requirements. • The technical and organizational measures for safety guarantee shall meet well-known results of the research investigations or shall be experimental validate

  6. Hybrid BN-600 core benchmark analyses

    International Nuclear Information System (INIS)

    Cross section library KAFAX used for BN-600 core benchmark calculations was based on nuclear data files ENDF-B/VI and JEF-2.2. Generation of effective cross sections were generated by a homogeneous cell model, group collapsing from 80 to 9 groups. Core neutron flux calculations were done by coarse mesh nodal diffusion approximation (DIF3D code), Nodal simplified P2 transport calculation (SOLTRAN code), and discrete SN approximation (TWODAT code). First order perturbation theory was used for reactivity parameter calculation. Calculation code applied for burnup calculation was the three-dimensional code REBUS-3 with 9 group cross section library from basic neutronic calculations. Results obtained include: multiplication factor k-eff at the beginning and at the end of cycled, reactivity burnup loss, fuel Doppler coefficient and sodium density coefficient. Results of heterogeneity calculations include k-eff, control rod worth and sodium density coefficient

  7. Benchmark analyses for BN-600 MOX core with minor actinides

    International Nuclear Information System (INIS)

    In 1999 the IAEA has initiated a Co-ordinated Research Project on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects'. Three benchmark models representing different modifications of the BN-600 reactor UOX core have been sequentially established and analyzed, including a hybrid UOX/MOX core, a full MOX core with weapons-grade plutonium and a MOX core with plutonium and minor actinides coming from spent LWR fuel. The paper describes studies for the latter MOX core model. The benchmark results include core criticality at the beginning and end of the equilibrium fuel cycle, kinetics parameters, spatial distributions of power and reactivity coefficients obtained by employing different computation tools and nuclear data. Sensitivity studies were performed to better understand in particular the influence of variations in different nuclear data libraries on the computed results. Transient simulations were done to investigate consequences of employing a few different sets of power and reactivity distributions on the system behavior at the initial phase of ULOF. The obtained results are analyzed in the paper. (author)

  8. Reactivity coefficients in BN-600 core with minor actinides

    International Nuclear Information System (INIS)

    In 1999, the IAEA has initiated a Coordinated Research Project on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects.' Three benchmark models representing different modifications of the BN-600 fast reactor have been sequentially established and analyzed, including a hybrid core with highly enriched uranium oxide and MOX fuel, a full MOX core with weapons-grade plutonium, and a MOX core with plutonium and minor actinides coming from spent nuclear fuel. The paper describes studies for the latter MOX core model. The benchmark results include core criticality at the beginning and end of the equilibrium fuel cycle, kinetics parameters, spatial distributions of power, and reactivity coefficients obtained by employing different computation tools and nuclear data. Sensitivity studies were performed to better understand in particular the influence of variations in different nuclear data libraries on the computed results. Transient simulations were done to investigate the consequences of employing a few different sets of power and reactivity coefficient distributions on the system behavior. The obtained results are analyzed in the paper. (author)

  9. Monte Carlo simulation of BN-600 LMFR hybrid core

    International Nuclear Information System (INIS)

    The safe operation of a large fast reactor requires accurate estimation of power produced in different parts of the reactor core and blanket. MCNPX code was used to develop a model to simulate and study the whole core of a prototype LMFR hybrid core; the BN-600. In this model, the core is composed of eight radial zones (typical code model layout is illustrated) the first two inner zones are low enrichment zones (LEZ), followed by a medium enrichment zone (MEZ). In the forth zone is the mixed oxide zone (MOX) composed of (U,Pu)O2 fuel subassemblies, then the outer high enrichment fuel zone (HEZ). The rest of the core are two zones of steel shielding assemblies (SSA) and an outer radial reflector to enclose the whole core. There is also 19 shim and control rods (SHR), and 6 scram rods (SCR). The model also take into account the axial variation in geometry and composition, this is accomplished by dividing the core axially into eight different zones with a definite thickness and composition. Partial insertion of control assembly which distorts the reactor flux and fission rates distribution are simulated using the three dimensional model of the reactor core. The spectrum of neutron flux is divided into 23 energy groups. Through this work several parameters are analyzed including criticality, axial and radial power distributions at different zones of the reactor core and burnup analysis in a typical operating conditions of the reactor core. F4 tally was used to calculate the flux distribution in the core and FM4 card was used to calculate the power distribution which is normalized to a total power of 1470 Mw. The energy release per fission was fixed to 200 Mev, as suggested in the BN-600 benchmark details. The temperature variation inside every cell (assembly) were considered by using the 'TMP' card. All fuel cells are at a uniform temperature 1500 K and all structural and coolant isotopes are at a uniform temperature 600 K, and in our model we assign a cross section

  10. Demonstration experiment of 3 BN-600 MOX vibropac FAs irradiation for the excess weapons plutonium disposal

    International Nuclear Information System (INIS)

    The paper presents results of a demonstration experiment on conversion of 50 kg of weapon-grade plutonium in the form of metal ingots into granulated MOX-fuel to be used for manufacturing fuel pins and 3 fuel assemblies (FAs) for the fast power-generating reactor BN-600, irradiation parameters of these FAs and the data from post-irradiation examinations. It can be concluded from the PIE results that the 3FAs were successfully irradiated in BN-600 without any fuel pin failures. Therefore, disposition of weapon - grade plutonium with a weight of about 20kg was successfully done. This represents the first disposition of Russian surplus weapon - grade plutonium as an international cooperation (this experiment was performed in collaboration between RIAR and JNC). The possibility of using MOX vipac fuel as a method for weapon plutonium disposition is clearly shown. (author)

  11. Results of nuclear design accuracy evaluation on BN-600 hybrid core

    International Nuclear Information System (INIS)

    Nuclear design accuracy on the BN-600 hybrid core has been evaluated using the JNC's nuclear analysis system for FBR cores, by utilizing the critical experiment analysis results on BFS-62 configuration that had been obtained under JNC's efforts for Russian surplus weapons plutonium disposition. In the BN-600 hybrid core design, a part of the current UO2 fuel region is replaced by MOX fuel, and the peripheral blanket region by stainless steel reflectors, respectively. These changes were simulated in a series of critical experiment configurations (BFS-62-1 to 4). Based on the analysis results on both BFS-62 configurations and other fast reactor cores, nuclear design accuracy on the BN-600 hybrid core has been evaluated by applying both the group constant adjustment method and the bias method. Evaluated nuclear parameters include, the ciriticality, fission rate distribution, sodium void reactivity, control rod worth, burn-up reactivity loss, etc. It is concluded, by applying the group constant adjustment method, that the evaluated accuracy (uncertainty) of most of the nuclear parameters can be decreased to less than half of those based on the basic nuclear constant without reflecting any experimental data. The improvement was mainly achieved by reducing the covariance of the iron elastic cross section. This significant effect results from the feature of the BN-600 hybrid core, which has relatively larger power density, adopts U235 as the main fissile nuclide, and has the stainless steel reflector surrounding the fuel region. In addition, good consistency of analysis results between the BFS and other fast reactor cores is confirmed. Information obtained by BFS-62 experiment show significant contribution to the accuracy improvement. It is also found that the bias method shows less significant effects on the accuracy improvement than the group constant adjustment method. Furthermore, the bias method may degrade the accuracy for certain nuclear parameters that have large

  12. BN-600 fully MOX fuelled core benchmark analyses (Phase 4). Draft synthesis report - Revision 1

    International Nuclear Information System (INIS)

    A benchmark analysis of a BN-600 fully mixed oxide (MOX) fuelled core design with sodium plenum above the core has been performed as an extension to the study of the BN-600 hybrid uranium oxide (UOX)/MOX fuelled core carried out during 1999-2001. This work was carried out within the the IAEA sponsored Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects'. This benchmark analysis retains the general objective of the CRP which is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The scope of the benchmark is to reduce the uncertainties of safety relevant reactor physics parameter calculations of MOX fuelled fast reactors and hence to validate and improve data and methods involved in such analyses. In previous benchmark analyses of the BN-600 hybrid core that closely conforms to a traditional configuration, the comparative analyses showed that sufficient accuracy is achieved using the diffusion theory approximation, widely applied in fast reactor physics calculations. With the purpose of investigating a core configuration of full MOX fuel loading, a core model of the BN-600 type reactor, designed to reduce the sodium void effect by installing a sodium plenum above the core, was newly defined for the next benchmark study. The specifications and input data for the benchmark neutronics calculations were prepared by EPPE (Russia). The specifications given for the benchmark describe only a preliminary core model variant and represent only one conceptual approach to BN-600 full MOX core designs. The organizations participating in the BN-600 fully MOX fuelled core benchmark analysis are: ANL from the USA, CEA and SA from EU (France and the UK, respectively), CIAE from China, FZK/IKET from Germany, IGCAR from India, JNC from Japan, KAERI

  13. BN-600 hybrid core benchmark analyses (phases 1, 2 and 3) (draft synthesis report)

    International Nuclear Information System (INIS)

    This report presents the results of benchmark analyses for a hybrid UOX/MOX fuelled core of the BN-600 reactor. This benchmark was proposed during the first Research Co-ordination Meeting (RCM) of the Co-ordinated Research Project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects, which took place in Vienna on 24 - 26 November 1999. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. There has been no change in the view that energy production with breeding of fissile materials is the main goal of fast reactor development to ensure long-term fuel supply. However, before the breeding role of fast reactors is recognized economically, due to the increasingly available low-cost uranium from the 1980s onwards, the emphasis of fast reactor development shifted to incineration of stock-piled plutonium and partitioning and transmutation (P and T) of nuclear wastes to meet contemporary demands. Following a proposal of the Russian Federation, at the 32nd Annual Meeting of the International Working Group on Fast Reactors (IWG-FR), held in May 1999, a hybrid UOX/MOX (mixed oxide) fuelled BN-600 reactor core that has a combination of highly enriched uranium (HEU) and mixed oxide (MOX) assemblies in the core region, was chosen as a calculational model. Hence the benchmark clearly addresses the issues of weapons-grade plutonium for energy production in a mixed UOX/MOX fuelled core of the BN-600 reactor. The input data for the benchmark neutronics calculations have been prepared by OKBM and IPPE (Russia). The input data have been reviewed and modified in the first RCM of this CRP. The organizations participating in the BN-600 hybrid core benchmark analyses are: ANL from the USA, CEA and SA (its previous name was AEAT) from EU (France and the

  14. Post-irradiation examination of the BN-600 core assemblies. The second modification of the hot cell equipment

    International Nuclear Information System (INIS)

    The current state of methodological support to the post irradiation examination of the reactor assemblies and their components operability implemented in accordance with the requirements of the regulations related to nuclear safety of nuclear reactors is described. The methodology of the examination is based on the experience of those mass primary post irradiation examinations of the BN-600 assemblies that have been performed in the spent fuel cooling pond and in the hot laboratory of Beloyarsk-3 reactor. This paper presents the main results of the second modification of the process and experimental equipment of the hot cell of BN-600 reactor carried out after the year of 2000. Further development prospects of the on-site examination complex are discussed

  15. Benchmark analyses for BN-600 MOX core with minor actinides

    International Nuclear Information System (INIS)

    Full text: The IAEA has initiated in 1999 a Coordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects'. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides (MAs). For this purpose, three benchmark models representing different modifications of the BN-600 reactor UOX core have been sequentially established and analyzed,the benchmark specifications being provided by IPPE. The first benchmark model is a hybrid UOX/MOX core, with UOX fuel in the inner core part and MOX fuel in the outer one, the fresh MOX fuel containing depleted uranium and weapons grade plutonium. The second model is a full MOX core, similar MOX fuel composition being assumed; a sodium plenum being introduced above the core to improve the core safety. The third model is analyzed in the paper. The model represents a similar full MOX core, but with plutonium and MAs from 60 GWd/t LWR spent fuel after 50 years cooling (thus assuming a so-called homogeneous recycling of MAs in a fast system). This option is the most challenging one (compared to those analyzed earlier in the CRP) as concerns the reactor safety since an increased content of MAs, in particular americium, and higher (than Pu239) isotopes of Pu leads to less favourable safety parameters. On the other hand, existing uncertainties in nuclear data for MAs and higher Pu isotopes may lead to relatively high uncertainties in the computation results for the considered model. The benchmark results include core criticality at the beginning and end of the equilibrium fuel cycle, kinetics parameters, spatial distributions of power and reactivity coefficients provided by CRP participants and obtained by employing different computation models and nuclear data. Sensitivity studies were performed at

  16. BN-600 fully MOX fuelled core benchmark analyses (phase 4) (draft synthesis report)

    International Nuclear Information System (INIS)

    This report presents the results of benchmark analysis for a fully Mixed Oxide (MOX) fuelled core of the BN-600 reactor. This benchmark analysis is an extension to the study of a hybrid UOX/MOX fuelled core performed during 1999 - 2001. These benchmark core analyses have been performed within the frame of the IAEA sponsored Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects' commenced in 1999. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. In the hybrid BN-600 core benchmark analyses, the substantial spread between the different participants noticed for several reactivity coefficients and power distributions did not have a significant impact on the transient behavior prediction, especially up to the onset of sodium boiling in the ULOF transient analyses. This result highlighted the compensating effects between several reactivity effects in the specific design of the hybrid core mainly loaded with UOX fuel. This gave confidence that the outcome of this type of transient could be understood in the partially MOX fuelled hybrid core type. From the recognition of significant interest of the analysis of a fully fuelled MOX core, a study of a BN-600 fully MOX fuelled core design with sodium plenum above the core including transient analyses has been defined for Phase 4 as an additional phase work in the 3rd Research Co-ordination Meeting (RCM) of the CRP. The specifications and input data for the benchmark neutronics calculations have been prepared by IPPE (Russia) and posted to the collaborator web site. The specifications given here describe a preliminary core model variant. It represents only conceptual approaches to a BN-600 full MOX core design and does not represent the real technical

  17. JNC results of BN-600 benchmark calculation (phase 3)

    International Nuclear Information System (INIS)

    The present work is the result of phase 3 BN-600 core benchmark problem, meaning burnup and heterogeneity. Analytical method applied consisted of: JENDL-3.2 nuclear data library, group constants (70 group, ABBN type self shielding transport factors), heterogeneous cell model for fuel and control rod, basic diffusion calculation (CITATION code), transport theory and mesh size correction (NSHEX code based on SN transport nodal method developed by JNC). Burnup and heterogeneity calculation results are presented obtained by applying both diffusion and transport approach for beginning and end of cycle

  18. The results of BN600 hybrid benchmark core analysis

    International Nuclear Information System (INIS)

    The present paper includes the results of phase 1 (RZ, two dimensional model) calculations of the BN-600 hybrid core benchmark problem. The methods applied consisted of: diffusion approximation with calculation of direct and adjoint problems; calculation of reactivity coefficients; ABBN-93 nuclear data library processing, (18 group calculations). Phase 2 (Hex-Z, three dimensional model) consists of diffusion approximation with calculation of direct and adjoint problems; calculation of reactivity coefficients using first order perturbation theory; nuclear data processing code for the ABBN-78 data library. Results presented include: multiplication factors, Doppler coefficients, fuel and structure density coefficients, expansion coefficients, power distribution, beta-effective values, reaction rate distributions

  19. Implementation of technology for reliability, safety and risk monitoring on BN-600 power unit

    International Nuclear Information System (INIS)

    For assessment of current safety level of Beloyarsk-3 reactor by probabilistic methods and for risk-informed decision supports during operation the reliability, safety and risk monitoring system of nuclear unit is developed and is putting into practice. The base of this system is the risk monitoring system RIM, which uses logico-probabilistic models of nuclear unit improved for risk monitoring aids. Risk monitoring system RIM is interconnected with maintenance and repair planning system of Beloyarsk-3 reactor and analytical monitoring system of reliability and safety. The analytical monitoring system of reliability and safety of BN-600 reactor is developed and adopted on the base of data retrieval system Istochnik-BN which provides structured storage of information, data systematization and processing, it has extensive data retrieval and exchange system

  20. Design study on BN-600 hybrid core. 1. Evaluation of core neutronic and thermalhydraulic characteristics by Japanese analysis methods

    International Nuclear Information System (INIS)

    A program of disposition of Russian weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600 hybrid reactor core has been progressed. The relevant design works on the BN-600 hybrid core were carried out under the contract between Japan Nuclear Cycle Development Institute (JNC) and OKB Mechanical Engineering (OKBM), Russian public enterprise, and completed. JNC obtained a series of design technical reports. Japanese analysis methods were adopted to evaluate neutronic and thermal-hydraulic characteristics of the BN-600 hybrid core, based on the design technical data described in the obtained reports. The evaluation results of the key core performances, such as maximum linear heat rate, core-averaged discharge burnup, sodium void reactivity, capability of disposition of weapon-grade plutonium and, and reactivity control balance, were found to satisfy the design criteria and/or targets provided by Russia, and meet the Russian rule. The results of this study showed that the core and fuel specifications determined by Russia can be considered reasonable and proper in terms of neutronic and thermal-hydraulic design, and that the Japanese analysis methods are expected to contribute to increasing reliability of the Russian design works. (author)

  1. Design study on BN-600 hybrid core. 2. Evaluation of fuel integrity and core neutronic characteristics by Japanese analysis methods

    International Nuclear Information System (INIS)

    A program of disposal of Russian surplus weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600 hybrid reactor core has been progressed. The relevant design works on the BN-600 hybrid core have been carried out under the contract between Japan Nuclear Cycle Development Institute (JNC) and OKB Mechanical Engineering (OKBM), Russian public enterprise. JNC obtained a series of design technical reports. Japanese analysis methods were adopted to evaluate fuel integrity in the design basis transients and neutronic characteristics of the BN-600 hybrid core, based on the design technical data described in the obtained reports. The evaluation results of the key performances, such as maximum cladding and fuel temperatures, coolant (sodium) void reactivity, reactivity coefficient, were found to satisfy the design criteria and/or target provided by Russia, and meet the Russian rule. The results of this study showed that the core and fuel specifications determined by Russia can be considered reasonable and proper from the viewpoint of safety and neutronic designs, and that the Japanese analysis methods are expected to contribute to increasing reliability of the Russian design works. (author)

  2. JNC results of BN-600 hybrid core benchmark calculations (3-D)

    International Nuclear Information System (INIS)

    This paper presents the phase 2 calculation results of the benchmark core of the BN-600 reactor (3-d modelling). The analytical method applied included the following: JENDL-3.2 nuclear data library; 70 group ABBN type self shielding factor table for group constants; reference delayed neutron yield and spectrum adopted; effective cross section obtained by SLAROM code; basic calculation done by using 18 group two dimensional RZ model (CITATION code) with region dependent fission spectra; transport theory and mesh size correction (TWOTRAN code); perturbation calculation done by diffusion, first order perturbation reactivity mapping method (PERKY code). Calculation results include effective multiplication factors; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution; beta and neutron life time; reaction rate distribution

  3. BN-600 hybrid core benchmark analyses (phases 1 and 2) (draft synthesis report)

    International Nuclear Information System (INIS)

    Tables of detailed calculation results submitted by eight participants (namely ANL, CEA/SA, CIAE, IGCAR, IPPE, JNC, KAERI, OKBM) of the CRP contain the following data: multiplication factor; Doppler coefficients for fuel and steel; reactivity density coefficient for sodium, steel and fuel; reactivity worth of absorption elements; expansion coefficients; power distribution in the core; reaction rate distributions. All the calculations were done for the chosen benchmark BN-600 reactor core. The report contains short description of the applied codes and nuclear data libraries used. The computer codes used for the cell calculations, i.e. calculations of effective cross sections applied Monte Carlo ultra fine method; subgroup method; f-factor method; collision probability method. Codes used for core calculations applied diffusion theory, transport theory, perturbation theory, Monte Carlo method using different condensed group constant sets

  4. Post-irradiation examination of the BN-600 reactor pins

    International Nuclear Information System (INIS)

    The diameter changes, microstructure, chemical attack of cladding by fission products and mechanical properties of pin cladding manufactured from the OX16H15M36 (C-0.09 wt.% Cr-16 wt.%, Ni-15 wt.%. Mo-3 wt.%. Nb-0.6 wt.%) 20% cold-worked stainless steel have been studied. It is shown that the main factors limiting the pin lifetime are large void swelling of the cladding and the mechanical interaction of the pin bundle with the wrapper. (author)

  5. BN-600 phase 4 (full MOX) reactivity coefficients and powers - hexagonal benchmark configuration

    International Nuclear Information System (INIS)

    The U.S. interest specific to benchmark configuration is: BN-600 hybrid core (Phase 2) and Full MOX core (Phase 4) option for joint U.S./Russia program; Interactions with Russian experts regarding safety analysis results and techniques; Accuracy of reactivity feedbacks is a key issue; Fast reactor analysis methods developed at ANL; Extensively tested in ZPR and ZPPR experiments; Applied to EBR-2 and FFTF test reactors; Safety analysis methods (and reactivity coefficients) used for analysis of passive safety tests in both systems. Calculations performed with demonstrated ANL suite of fast reactor analysis tools. Complete set of results generated for Phase 4 and Phase 2 includes: Basic parameters; Fuel and structure regional Doppler coefficients; Regional material worths; Kinetics parameters. In general, whole core Phase 4 (Full MOX) and Phase 2 (Partial MOX) results are similar. Only exception is sodium worth. Benchmark results should enable detailed (regional) comparison of key reactivity feedbacks. How will results be used to clarify uncertainties? Sodium Void Worth Calculation were performed for benchmark calculations using homogeneous diffusion theory. For heterogeneous effects, see 'Evaluation of Benchmark Calculations on a Fast Power Reactor Core with Near-Zero Sodium Void Effect', IAEATECDOC-731. A very important result has been achieved in the calculation of reaction rate distribution for configuration with reflector in direct contact with the core (no presence of blanket). The use of a very large number of groups (∼1000) has allowed to accurately reproduce the spectrum transient and consequently dramatically improve the results. This solves a longstanding (more than a decade) discrepancy for these kinds of configurations. Following these findings an iterative methodology, based on conservation of reaction rates, has been successfully developed for allowing to reproduce the same type of reaction rate distributions obtained with the large of number of

  6. JNC's review and proposal for BN-600 hybrid core benchmark calculation

    International Nuclear Information System (INIS)

    This contribution includes questions on benchmark description (geometry, composition, data evaluation) and proposals for the BN-600 benchmark project. Proposals are related to benchmark of cell heterogeneity evaluation (fuel assembly, control rod); additional burnup properties (burnup reactivity loss, fuel composition change); analysis by using the cross section sensitivity method (application of perturbation theory, influence of cross section difference, estimation of analytical method difference); evaluation of BN-600 design value and its errors (best estimated design value of hybrid core, error estimation of the design value)

  7. Modernization of RTC for fabrication of MOX fuel, Vibropac fuel pins and BN-600 FA with weapon grade plutonium

    International Nuclear Information System (INIS)

    Since mid 70's RIAR has been performing activities on plutonium involvement in fuel cycle. These activities are considered a stage within the framework of the closed fuel cycle development. Developed at RIAR fuel cycle is based on two technologies: 'dry' process of fuel reprocessing and vibro-packing method for fuel pin fabrication. Due to the available scientific capabilities and a gained experience in operating the technological facilities (ORYOL, SIC) for plutonium (various grade) blending into fuel for fast reactors, RIAR is a participant of the activities aimed at solving these tasks. Under international program RIAR with financial support of JNC (Japan) is modernizing the facility for granulated fuel production, vibro-pac fuel pins and FA fabrication to provide the BN-600 'hybrid' core. In order to provide 'hybrid' core it is necessary to produce (per year): - 1775 kg of granulated MOX-fuel, 6500 fuel pins, 50 fuel assemblies. Potential output of the facility under construction is as follows: - 1800 kg of granulated MOX-fuel per year, 40 fuel pins per shift, 200 FAs for the BN-600 reactor per year. Taking into account domestic and foreign experience in MOX-fuel production, different options were discussed of the equipment layouts in the available premises of chemical technological division of RIAR: - in the shielded manipulator boxes, in the existing hot cells. During construction of the facility in the building under operation the following requirements should be met: - facility must meet all standards and regulations set for nuclear facilities, installation work at the facility must not influence other production programs implemented in the building, engineering supply lines of the facility must be connected to the existing service lines of the building, cost of the activities must not exceed amount of JNC funding. The paper presents results of comparison between two options of the process equipment layout: in boxes and hot cells. This equipment is intended

  8. Introduction of reliability, safety and risk monitoring technology in BN-600 power unit

    International Nuclear Information System (INIS)

    Rosenergoatom Concern OJSC Operating Entity set the task to introduce risk monitoring technology for continuous estimate and control of nuclear units quantitative safety measures change. JSC “Afrikantov OKBM” and “Beloyarsk NPP” developed risk monitor system “RIM” which is introducing now at Beloyarsk NPP BN-600 unit 3. To estimate quantitative safety measures Level 1 PSA for internal initiating events for full power operating conditions model of Beloyarsk NPP BN-600 unit 3 is used. PSA model was developed using national certified PSA software CRISS. To ensure NPP reliability and safety, implementation of comprehensive systematic study (monitoring) of NPP operating experience is of fundamental value. To solve this problem, JSC “Afrikantov OKBM” and “Beloyarsk NPP” developed and introduced the system for analytical reliability and safety monitoring of BN-600 power unit based on information retrieval system (IRS) “Istochnik-BN”. The paper describes system objectives, main characteristics and results of reliability, safety and risk monitoring technology introducing at Beloyarsk NPP BN-600 unit 3. (author)

  9. Preliminary analysis of the proposed BN-600 benchmark core

    International Nuclear Information System (INIS)

    The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

  10. BN-600 hybrid core benchmark analyses. Results from a coordinated research project on updated codes and methods to reduce the calculational uncertainties of the LMFR reactivity effects

    International Nuclear Information System (INIS)

    To those Member States who have or have had significant fast reactor development programmes, it is of the utmost importance to have validated up-to-date codes and methods for fast reactor core physics analysis in support of R and D activities in the area of actinide utilization and incineration. They have recently focused on fast reactor systems for minor actinide transmutation and on cores optimized for consuming rather than breeding plutonium; the physics of the breeder reactor cycle having already been widely investigated. Plutonium burning systems may have an important role in managing plutonium stocks until the time when major programmes of self-sufficient fast breeder reactors are established. For assessing the safety of these systems it is important to determine the prediction accuracy of transient simulations and their associated reactivity coefficients. In response to Member States' expressed interest, the IAEA sponsored a Coordinated Research Project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. This CRP was started in November 1999 and at the first meeting the members of the CRP endorsed a benchmark on the BN-600 hybrid core for consideration in its first studies. Benchmark analyses of the BN-600 hybrid core were performed during the first three phases of the CRP investigating different nuclear data and levels of approximations in the calculation of, safety related reactivity effects and their influence on uncertainties in transient analysis predictions. In an additional phase of the benchmark studies experimental data was used for the validation and verification of nuclear data libraries and methods in support of the previous three phases. This report presents the results of the benchmark analyses of the hybrid UOX/MOX fuelled BN-600 reactor core. The aim of this report is to contribute to the reduction in uncertainties associated with reactivity coefficients and their influence on LMFR

  11. BN-600 MOX Core Benchmark Analysis. Results from Phases 4 and 6 of a Coordinated Research Project on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects

    International Nuclear Information System (INIS)

    For those Member States that have or have had significant fast reactor development programmes, it is of utmost importance that they have validated up to date codes and methods for fast reactor physics analysis in support of R and D and core design activities in the area of actinide utilization and incineration. In particular, some Member States have recently focused on fast reactor systems for minor actinide transmutation and on cores optimized for consuming rather than breeding plutonium; the physics of the breeder reactor cycle having already been widely investigated. Plutonium burning systems may have an important role in managing plutonium stocks until the time when major programmes of self-sufficient fast breeder reactors are established. For assessing the safety of these systems, it is important to determine the prediction accuracy of transient simulations and their associated reactivity coefficients. In response to Member States' expressed interest, the IAEA sponsored a coordinated research project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. The CRP started in November 1999 and, at the first meeting, the members of the CRP endorsed a benchmark on the BN-600 hybrid core for consideration in its first studies. Benchmark analyses of the BN-600 hybrid core were performed during the first three phases of the CRP, investigating different nuclear data and levels of approximation in the calculation of safety related reactivity effects and their influence on uncertainties in transient analysis prediction. In an additional phase of the benchmark studies, experimental data were used for the verification and validation of nuclear data libraries and methods in support of the previous three phases. The results of phases 1, 2, 3 and 5 of the CRP are reported in IAEA-TECDOC-1623, BN-600 Hybrid Core Benchmark Analyses, Results from a Coordinated Research Project on Updated Codes and Methods to Reduce the

  12. Neutronic safety parameters of the BN-600 type reactor with hybrid core. Draft

    International Nuclear Information System (INIS)

    Diffusion method in a standard 26 group approximation was used for the initial condition when calculating reactivity coefficients. Space distribution of the reactivity coefficients was obtained using first order disturbance theory (their integral values were normalised to the relative direct calculation results). This methodology is justified by both quite acceptable accuracy of diffusion calculation and comparability to the similar results obtained. In this connection, results presented below realise the following approximations: RHEIN (26 groups diffusion theory, R-Z, homogeneous media); TRIGEX (18 groups diffusion theory, GEX-Z, homogeneous media). Three main directions were taken for studies: definition of possible error of diffusion approach in flux calculation; influence of model simplifying; calculation of reactivity effects (integral value and space distribution) using traditional (diffusion) methods. (author)

  13. The influence of an energy groups number and mesh size on results of reactivity coefficients calculations for BN-600 benchmark core. Appendix 1

    International Nuclear Information System (INIS)

    This document presents the OKBM contribution to the analysis of a benchmark of BN-600 reactor hybrid core with simultaneous loading of uranium fuel and MOX fuel within the framework of the international IAEA Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects'. The purpose of the present document is the comparison of some obtained for the Phase 2 results using the different energy groups number and the different mesh point size. The CRP participants at calculation of a benchmark used different planar mesh point sizes. The axial mesh point size was not stipulated, but the mesh sizes were specified for the desired results representation. Therefore in some cases there was possible the application of rather large axial mesh size. The discrepancy in results because the different mesh point size using should be estimated. The results of some participants were obtained using the relatively small energy groups number - 6, 9, and 12. The influence of the energy group number on value of the obtained reactivity coefficients is analyzed in case of the OKBM calculations results. Besides in case of the sodium density reactivity coefficient the OKBM used method of the choice for the few group optimal division of the energy scale is shown. The probable additional uncertainties as the consequence of the baseless group division are estimated by the comparison of the group division schemes applied by different CRP participants

  14. Comparison of non-leakage and leakage terms of SVRW in BFS-62-3A and BN-600 hybrid core. Appendix 9

    International Nuclear Information System (INIS)

    As was presented in PHYSOR2004, there exists a difference in region-wise SVRW values between BFS-62-3A and BN-600 hybrid core. Here shows the breakdown into non-leakage and leakage terms for both the cores. Please note that SVRW in BN-600 hybrid core was calculated by voiding all fuel assemblies in each zone and subsequently divided by 6 to obtain an equivalent value to those in BFS-62-3A. Another point to be noted is that all the results were based on heterogeneous cell calculation in fuel zones, which is different from the condition discussed in IAEA CRP. One item to be noted is relatively negative SVRW measured in BFS-62-3A than that calculated in the BN-600 hybrid core, as shown. The significant difference in SVRW is mainly due to temperature and CR axial position differences. It surely suggests both difficulty in applying the bias correction method and possible effectiveness of the nuclear constant adjustment method, in decreasing predicted uncertainty of SVRW of the BN-600 hybrid core by reflecting BFS-62-3A measurement data

  15. Possible remaining reasons of the differences on SVRW between BFS-62-3A and BN-600 hybrid core. Appendix 8

    International Nuclear Information System (INIS)

    The document Action 5.12v1 prepared by Dr. G.Rimpault mentions that there are unknown reasons on the differences of SVRW between the BN-600 hybrid core and BFS-62-3A. In this note possible reasons are investigated for LEZ and MOX region voiding. Calculations were based on heterogeneous cell calculation in fuel zones, and SVRW in BN-600 hybrid core was calculated by voiding all fuel assemblies in each zone and subsequently divided by 6 to obtain equivalent values to those in BFS-62-3A. Effects of the following parameters on SVRW were investigated. 1. Na density difference; 2. Temperature difference; 3. Pu presence in LEZ, MEZ, and HEZ (1.0∼1.7wt% only in BN-600); 4. FP presence (only in BN-600); 5. Control rod partially inserted (only in BN-600). The investigation was made in a step-by-step manner by removing the differences from BN-600 hybrid core model, except that the difference in Na density was removed by changing BFS-62-3A results. Since what is of interest in IAEA CRP Phase 3 (BN-600 hybrid core benchmark model analysis) is the sodium density coefficient, the difference was adjusted at first. The difference of the 3rd parameters was roughly removed by changing Pu in the UO2 regions to U-235. Obtained reactivity data are compared in a form of upper, lower, and total energy integrated values with a energy boundary of 31.8 keV, since the most influential scattering (slowing down) components has change sign around 30-40 keV. The effect of the 5th parameter is found to be negligible and not shown. The scattering components are compared in the 70 energy group structure, for LEZ and MOX, respectively. It is clearly found in LEZ region case that the differences are reduced as steps proceed. The 3rd parameter affects the higher energy region. In MOX region, temperature effect is found but the other 2 parameters are not clear. The other differences such as Pu concentration in the region may affect the remaining difference

  16. Analysis of the BN-600 fast-spectrum core mock-up at BFS-2 zero-power facility using MCNPX

    International Nuclear Information System (INIS)

    Highlights: ► We model the BFS-62-3A experiment with the MCNPX code and four nuclear libraries. ► We show the impact on reactivity of heterogeneous structures in the reactor. ► We model experimental uncertainties, e.g. in materials dimension and density. ► The model agrees with experiments on k-eff, CR worth, Na voids and fission rates. ► The analysis questions experimental data measured in the reflector region. - Abstract: A 3D full-core heterogeneous model of the BFS-62-3A critical benchmark experiment was developed and validated using the Monte Carlo MCNPX-2.4.0 code. The BFS-2 critical facility at the Institute of Physics and Power Engineering (IPPE) was designed for simulation of fast reactor core neutronics, and for the validation of codes and nuclear data. The BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor core with (U, Pu) O2 fuel of 17% Pu content and stainless-steel reflectors. It was operated to measure the effective multiplication factor, spectral indices, radial fission rate distributions, control rod worths and sodium void effects. In the present study, special care was taken to run the MCNPX model to make Monte-Carlo confidence intervals comparable with uncertainties reported in the experiments; such as in material dimensions, number densities and isotopic compositions. In addition to the effective multiplication factor, sodium void effect, fission rate distributions and control rod worth were calculated. Simulations were carried out with four different modern nuclear data libraries; the primary aim being to estimate sensitivity of the results to the nuclear data. This task, besides being a library comparison, is also meant as a first step towards a code-to-code verification with deterministic methods. Results agree well with experimental values on most of the nuclear characteristics, even though a discrepancy up to more than 20% was found on the flux distribution in the stainless-steel reflector

  17. Analysis of the dynamics of hydrogen ingress to secondary sodium after the replacement of the steam generator stages of the BN-600 power unit

    International Nuclear Information System (INIS)

    The article presents the calculation methodology of hydrogen ingress rate into secondary circuit sodium of the BN-600 power unit. The sources and the mechanism of the hydrogen ingress after the replacement of a lot of the PGN-200M steam generator stages within the scope of work on the power unit operation lifetime extension were defined. The hydrogen ingress rate after impending replacements of steam generator stages was estimated

  18. Fast reactor operating experience gained in Russia: Analysis of anomalies and abnormal operation cases

    International Nuclear Information System (INIS)

    Review of various anomalous events and abnormal operation experience gained in the process of Russian fast reactors operation is given in the paper. The main information refers to the BN-600 demonstration reactor operation. Statistical data on sodium leaks and steam generator failures are presented, and sources of these events and countermeasures taken to avoid their appearance on the operating reactors as well as related changes made in the BN-800 reactor design are considered. In the paper, some features of impurities behaviour are considered in various modes of the BN-600 reactor operation. Information is given on the impurities ingress into the circuits, on abnormal situation emerged in the process of the BN-600 reactor operation and its probable cause. Information is presented on the event related to the increased torque of the BN-600 reactor central rotating column and repair works performed. (author)

  19. Studies of residual heat release of the BN-600 reactor spent assemblies in a fuel cooling pond

    International Nuclear Information System (INIS)

    A method was developed, and a facility was manufactured for measuring residual heat release of spent fuel assemblies directly in the fuel pool. The operations involving rearrangement of spent fuel assemblies are performed using the standard hardware and technologies, which is the main advantage of the method. Thus, the design safety of handling operations is provided. Direct measurements of residual heat release of essential number of spent fuel assemblies of diverse types have been taken for the first time

  20. Operational capability of fuel elements with austenitic stainless steel cans i radiated in BN-600 reactor up to high burnup

    International Nuclear Information System (INIS)

    Aimed to reveal the factors limiting operational capability of fuel elements a study is made into the behaviour of ChS-68 and EhP-172 cold worked steel fuel cans irradiated up to damaging doses of 61.3-87.5 dpa. Fuel cans are shown to fail due to degradation of mechanical properties and subsequent microcrack formation at internal surface in a zone of maximal increase of diameter

  1. The analysis of behaviour of pin claddings made of 16Cr15Ni3MoNbB steel at high fuel burn-ups in BN-600

    International Nuclear Information System (INIS)

    Postreactor investigations into BN-600 fuel cans made of the 0Cr16Ni15Mo3NbB steel after irradiation to maximum burnup 10% heav. at. and higher have been conducted. It is demonstrated that the most degradation of performance attributes takes place in the zone of maximum increasing diameter of fuel cans. Processes leading to the degradation of fuel cans (embrittlement, corrosion resistance lowering) are connected either with swelling or radiation-induced segregation being in the same temperature range and under the same motive force that swelling is

  2. 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

  3. Operating experience of fast breeder reactors in the USSR

    International Nuclear Information System (INIS)

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

  4. 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)

  5. Modernization of OIK unit for fabrication of mixed-oxide fuel, vibrocompacted fuel elements and fuel-containing assemblies of BN-600 reactor using plutonium of weapon quality

    International Nuclear Information System (INIS)

    In the framework of participation in international project of weapon plutonium utilization modernization of the technological complex for fabrication of granulated fuel, vibrocompacted fuel elements and fuel-containing assemblies is realizing. Taking into account domestic and foreign experience of MOX-fuel fabrication different versions of equipment are examined

  6. Testing experimental fuel elements of the BN-600 fuel element type up to various depth of burn up in the BOR-60 reactor

    International Nuclear Information System (INIS)

    Results of the investigation of experimental fuel elements are presented. The authors discuss fuel element construction, basic testing parameters, results of measuring gas release from fuel, deformation of cladding and swelling of steel, and also data on material investigations of macro- an micro-structures of fuel and cladding with an analysis of the degree and character of their physico-chemical interaction with fission fragments

  7. Validation of SOCRAT-BN code on the base of reactor experiments

    International Nuclear Information System (INIS)

    SOCRAT-BN code is developed for safety assessment for Liquid Metal Fast Breeder Reactors (LMFBR) with sodium coolant during the design basis accidents and beyond design-basis accidents. Code consists of the several modules, which allows simulating: thermal hydraulic, neutron-physic, thermal-mechanical processes and behavior of fission products in rector vessel in coupled statement. Brief description of SOCRAT-BN code is presented in the paper. Validation on the base of the integral experiments and out-of-pile experiments were carried out to show code workability. Integral tests include experiments provided at BN-600 reactor. Detailed nodalization scheme, consists of primary, secondary and third sides, was performed to simulate BN-600 reactor. To validate SOCRAT-BN code the following BN-600 transients were chosen: natural circulation core cooling at 50% power load, and one loop tripping at about 100% power load. SOCRAT-BN code simulations results showed good agreement with the experimental data. (author)

  8. Some questions of fast reactor residual heat removal using coolant natural circulation in the circuits

    International Nuclear Information System (INIS)

    A calculation procedure designed for entire plant dynamics has been used for transients leading the reactor plant of BN-600 type into a natural circulation mode. A detailed description of the mathematical model and comparison of calculated and experimental results are presented. Some peculiarities of coolant natural circulation development in the BN-350 reactor circuits are considered. (author)

  9. 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)

  10. 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

  11. 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)

  12. 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)

  13. 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)

  14. An analysis of fast reactor fuel assembly performance taking into account their mechanical interaction in the core and refuelling line capabilities

    International Nuclear Information System (INIS)

    An approach to assessment of fast reactor fuel assembly performance has been considered. A concept of passive restraint of fuel assemblies in a reactor adopted in the USSR is described. Some methods for calculating the interassembly interactions during operation are briefly outlined, some calculated results are presented. A problem of fuel assembly performance during refuelling taking into account the refuelling line capabilities is considered. Some results from fuel assemblies operation experience in the BN-600 reactor are given. (author)

  15. 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

  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. Assessment of the dry process fuel sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

  18. The development of fast neutron power reactors with a sodium coolant and ways in which their technical and economic performance can be improved

    International Nuclear Information System (INIS)

    During the years that have elapsed since the commissioning of power units with fast neutron BN-350 reactors (1973) and BN-600 reactors (1980), considerable experience has been acquired in the operation of fuel elements, sodium equipment and mechanisms and steam generators. Operating experience has confirmed the reliability and safety of both types of facilities. Nevertheless, during the last few years a number of improvements have been introduced in the operating mode, in the equipment design solutions and in the system flow sheets. These solutions were aimed at further increasing the reliability, safety and profitability of the power units. The BN-800 reactor is based to a considerable extent on the scientific and technical ideas and design development of the BN-600 reactor. During its design, considerable attention was paid to developing reliable equipment which would ensure a higher level of fuel burnup than that of the BN-600 and to mastering the closed fuel cycle. The BN-1600 reactor constitutes a new step in the development and creation of a fast breeder reactor with sodium coolant. However, even in its design, the basic scientific and technical ideas developed for the BN-600 and BN-800 reactors were maintained to a considerable degree. The experience of developing and operating fast sodium cooled reactors in the USSR has confirmed the expected positive safety characteristics of this type of reactor. At the same time, there are significant reserves for further improving the safety characteristics and this does not present any major difficulties. (author). 5 refs, 1 fig., 4 tabs

  19. 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

  20. Experience Gained in the Russian Federation on Sodium Cooled Fast Reactors and Prospects for their Further Development

    International Nuclear Information System (INIS)

    Experience gained with sodium cooled fast reactors (SFRs) in the Russian Federation over the past 30 years is reviewed. Some statistical data on the operation indicators gained for SFRs worldwide and in the Russian Federation are presented. The basic phases of SFR technology development in the Russian Federation are described. The main research work undertaken on SFRs in the Russian Federation (BR-5/BR-10, BOR-60, BN-600) for justification of SFR technology is highlighted. Priority is given to analysis of operational experience of industrial power unit No. 3 of the BN-600 reactor at the Beloyarsk nuclear power plant and the operation indicators achieved. Statistical information is presented on abnormal events that occurred during operation of the BOR-60 and BN-600 units, and the extent of their influence on the facilities' safety, technical and economic performances. Conclusions on the level of mastery of SFR technology achieved in the Russian Federation are made, based on the review of operation experience, and prospects for their further development are estimated in the light of this experience. (author)

  1. Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

    OpenAIRE

    Darnowski Piotr; Uzunow Nikolaj

    2015-01-01

    An analysis of the influence of addition of minor actinides (MA) to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code...

  2. Experience gained in Russia on sodium cooled fast reactors and prospects of their further development

    International Nuclear Information System (INIS)

    Full text: The experience on sodium cooled fast reactors (SFR) gained in Russia in recent 20-30 years has been reviewed, and prospects of their further evolution have been analyzed taking into account the said experience. Consistent strategy of Russia in the area of fast reactors in recent 50 years has given grounds for gaining the world's leading positions in SFR. The experience accumulated on SFR operation is especially impressive. It covers research reactors BR-5/10 (Obninsk) and BOR-60 (Dimitrovgrad), demofacility BN-350 which is now in the territory of Kazakhstan Republic (Aktau), and the first commercial power unit with fast reactor BN-600 (Zarechny). Total experience with SFR operation in the USSR and in Russia amounts to over 140 reactor years (1/3 of entire world experience of operation). The construction of BN-800 reactor and the development of advanced commercial large-size SFR design has been underway (BN-K). The analysis of the experience with operation of BN-600 and BOR-60 reactors has been highlighted. Statistical information has been presented on the operation indicators of the commercial power unit with BN-600 reactor, with data for recent years covered (load factor, energy production, service indicators achieved for the main elements of the reactor facility (RF), including the fuel burn-up level mastered). The program of systematic increase of fuel burn- up realized on the BN-600 has been described with principal characteristics of the BN-600 core modifications realized in the process of RF operation.A steady functioning of the BN-600 reactor and high economic indicators achieved testify to a reliable mastering of the SFR technology, which covers technology of sodium coolant, operation and replacement of the main sodium equipment (steam generators, intermediate heat exchanger, MCP). A high level of safety has been demonstrated for SFR, with minimum effects on the personnel and environment, both in terms of gaseous releases, and volumes of liquid

  3. 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.

  4. Development of Fast Sodium Reactor Technology in the Russian Federation

    International Nuclear Information System (INIS)

    The paper provides information about the development of the sodium cooled fast reactors in the former USSR (Russian Federation) starting from the 1950s. The evolution of this technology is traced from the small research reactors to large power units. It is shown how power and parameters were changing in reactor plants; how engineering solutions on the layout, reactor core design, main equipment and systems were evolving; and how the most important problem on increasing the fuel burnup was being gradually solved. Mastering of the dense nitride fuel instead of the MOX fuel is mentioned as an important challenge. Given are operational results for the first power units with the BN-350 and BN-600 reactors; the experience obtained is evaluated. Characterized are the challenges to be faced in the new BN-800 and BN-1200 projects, as well as information about the status of these projects. (author)

  5. Approaches to validation of fast reactor lifetime extension

    International Nuclear Information System (INIS)

    As compared with the other reactors, the main feature of fast neutron sodium-cooled reactor operation is the effect of increased temperatures (up to 550-600 deg. C) and intensive fast neutron flux (up to ∼2·1021 n/cm2·year of energy E>0.1 MeV) on structural materials. Under these conditions, the basic mechanisms damaging fast reactor component material are creep, fatigue and their interaction as well. Under intensive neutron flux, the austenitic steel used as reactor structural material is embrittled. Except thermo-mechanical load, the additional loading factor for fast reactor components is non-uniform material swelling due to the effect of high-dose irradiation. This results in considerable deformation of reactor components that can violate their normal operation in the ultimate case. It is necessary to note that most of main fast reactor components are difficult of access for non-destructive inspection in order to detect defects. Therefore, in case of reactor service lifetime prolongation, it is necessary to take account of availability of process and operation defects in these components. In view of above-mentioned, to validate prolongation of BN-600 reactor service life up to 45 years, procedural and material-study activities were performed to develop procedures and methods of strength and lifetime analysis for structural components with defects under the effect of high temperatures and intensive irradiation, as well as to obtain service characteristics of fast reactor structural materials in view of their degradation under the effect of high temperature during more than 2.105 hours and intensive neutron irradiation. Within procedural tasks, the following main procedures and methods have been developed: 1. Definition of design dimensions and shapes of postulated defect. 2. Formulation of constitutive equations of thermal-viscoelastic-plastic deformation of structural material in view of swelling to analyze stress-strain state (SSS) of structural components

  6. After-heat analysis of BN-600 assemblies

    International Nuclear Information System (INIS)

    A method of after-heat calculation used in CARE-03 code and OST module of GEFEST package is described. The method was verified with ORIGEN code after-heat calculations. The calculation results validation against after-heat measurements with calorimetric equipment in Beloyarsk NPP fuel storage pool was carried out. The CARE code and ORIGEN code calculation results were shown to be in good agreement. The discrepancy of after-heat calculation data with measurement data is indicated for spent subassemblies under the long cooling times with low after-heat (100-150 W). The calculation results and measurement data agreement was less than 10% under short cooling times

  7. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-15

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology.

  8. Nuclear reactor fuel cycle technology with pyroelectrochemical processes

    International Nuclear Information System (INIS)

    A group of dry technologies and processes of vibro-packing granulated fuel in combination with unique properties of vibro-packed FEs make it possible to implement a new comprehensive approach to the fuel cycle with plutonium fuel. Testing of a big number of FEs with vibro-packed U-Pu oxide fuel in the BOR-60 reactor, successful testing of experimental FSAs in the BN-600 rector, reliable operation of the experimental and research complex facilities allow to make the conclusion about a real possibility to develop a safe, economically beneficial U-Pu fuel cycle based on the technologies enumerated above and to use both reactor-grade and weapon-grade plutonium in nuclear reactors with a reliable control and accounting system

  9. 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

  10. 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

  11. Design and layout decisions for refuelling system of advanced fast neutron reactor

    International Nuclear Information System (INIS)

    The experience in operation of BOR-60, BN-350 and BN-600 power units, as well as development of refuelling systems for BN-800 power unit, allows developing of refuelling system for BN-1200 advanced reactor of new generation. The refuelling system was developed on the basis of possible technical decisions aimed at improvement of safety and technical-and-economic indices. Structural layout of BN-1200 reactor refuelling system is given. Main differences in BN-1200 reactor refuelling system as compared with BN-800 reactor are given. Design features of refuelling equipment are: - BN-1200 reactor has a split large rotating plug to allow transporting of its components by railway with subsequent assembling at site; - the refuelling box is fabricated in the form of sectional parallelepiped to allow transporting of its components by railway with subsequent assembling at site; - one 'direct' refuelling mechanism and one cantilever' refuelling mechanism are used to refuel rarely replaced protection assemblies that allows reducing of overall dimensions of rotating plugs; - the vertical elevator is arranged on the oval plug installed on the reactor cover. The upper structure with elevator drive rotates together with the elevator plug under rotary drive located on the oval plug. The vertical elevator allows sufficient reduction of refuelling box; - the refuelling machine runs on straight-line rails. The vertical elevator, gas gate valve on reactor refuelling channel, non-use of spent FA drum and enhanced radiation protection on the column of refuelling box machine allows reduction of specific materials consumption of BN-1200 reactor refuelling system by more than 10 times as compared with BN-800 reactor. To verify refuelling equipment operability the following experiments are planned: - mastering of gripper design for 'direct' refuelling mechanism and refuelling machine; - mastering of 'cantilever' for refuelling mechanism; - mastering of fresh FA conveyor design. As for the

  12. Verification calculations as per CFD FLOWVISION code for sodium-cooled reactor plants

    International Nuclear Information System (INIS)

    The paper studies the experience in application of CFD FlowVision software for analytical validation of sodium-cooled fast reactor structure components and the results of performed verification, namely: – development and implementation of new model of turbulent heat transfer in liquid sodium (LMS) in FlowVision software and model verification based on thermohydraulic characteristics studied by experiment at TEFLU test facility; – simulation of flowing and mixing of coolant with different temperatures in the upper mixing chamber of fast neutron reactor through the example of BN-600 (comparison with the results obtained at the operating reactor). Based on the analysis of the results obtained, the efficiency of CFD codes application for the considered problems is shown, and the proposals for CFD codes verification development as applied to the advanced sodium-cooled fast reactor designs are stated. (author)

  13. Primary Damage Characteristics in Metals Under Irradiation in the Cores of Thermal and Fast Reactors

    International Nuclear Information System (INIS)

    For an analysis and forecasting of radiation-induced phenomena in structural materials of WWERs, PWRs and BN reactors the fast neutron fluence is usually used (for structural materials of the reactor cores and internals the fluence of neutrons with energy > 0.1 MeV, for WWER and PWRs vessel steels the fluence of neutrons with energy > 0.5 MeV in Russia and East Europe, and with energy > 1.0 MeV in USA and France). Displacements per atom (dpa) seem to be a more appropriate correlation parameter, because it allows comparing the results of materials irradiation in different neutron energy spectra or with different types of particles (neutrons, ions, fast electrons). Energy spectra of primary knocked atoms (PKA) and 'effective' dpa, which are introduced to take into account the point defect recombination during the relaxation stage of a displacement cascade, can be still better representation of the effect of irradiation on material properties. In this work the results of calculating dose rates (dpa/s, NRT-model), PKA energy spectra and PKA mean energies in metals under irradiation in the cores of Russian reactors WWER-440, WWER-1000 (both power thermal reactors) and BN-600 (power fast reactor) and BR-10 (test fast reactor) are presented. In all the reactors Fe and Zr are considered, with addition of Ti and W in BN-600. 'Effective' dose rates in these metals are calculated. Limitations and uncertainties in the standard dpa formulation (the NRT-dpa) are discussed. IPPE activities in the fields related to the TM subject are considered

  14. Estimation of accuracy for calculation of neutron field distribution in fast reactor on the reactor experiments basis

    International Nuclear Information System (INIS)

    Calculation of a neutron field distribution in fast reactor is the basic method for forecasting of such characteristics as local and average energy release values in pins and subassemblies, radiation doze on constructional materials, burning out of fuel. And calculation remains dominating means of the control over these parameters. Accuracy of calculation is numerical expression for reliability of definition of the specified characteristics and base for formation of design limits and reserves. The purpose of the given work consist in estimation of accuracy of modern computing means for calculation by the analysis of the available experimental data received on BFS critical facility (IPPE, Russia) and BN-600 power reactor. The comparison analysis of calculation and experimental data on measurement of fission rates distributions for 235U, 238U, 239Pu at critical assemblies BFS 62-th, BFS 66-th series and integrated quantity of fissions at BN-600 subassemblies by gamma-scanning method has been fulfilled. For carrying out of the analysis the integrated system MODEXSYS has been developed. This system includes up-to-date version of diffusion program TRIGEX, precision program MMKKENO, database of experimental data and benchmark models of experiments. Combination of various calculation methods allowed to reveal methodical uncertainties of calculation. Calculation of energy release distribution in traditional cores of fast reactors with sodium coolant is carried out on the average with accuracy 2-3%, the maximum discrepancies do not exceed 5%. On the border of core and radial blanket the methodical error of diffusion approximation, reaching 10% for 238U fission reaction rate, is experimentally confirmed. In the traditional uranium blanket an error reach value of 20%, and discrepancies have regular character: calculation data give lower values of reactions rates in relation to experiment. At transition to the steel reflector character of calculation-experimental discrepancies

  15. Main research results in the field of nuclear power engineering of the Nuclear Reactors and Thermal Physics Institute in 2014

    International Nuclear Information System (INIS)

    The main results of scientific and technological activities for last years of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in solving problems of nuclear power engineering are presented. The work have been carried out on the following problems: justification of research and development solutions and safety of NPPs with fast reactors of new generation with sodium (BN-1200, MBIR) and lead (BREST-OD-300) coolants, justification of safety of operating and advanced NPPs with WWER reactor facilities (WWER-1000, AEhS 2006, WWER-TOI), development and benchmarking of computational codes, research and development support of Beloyarsk-3 (BN-600) and Bilibino (BN-800) NPPs operation, decommissioning of AM and BR-10 research reactors, pilot scientific studies (WWER-SKD, ITER), international scientific and technical cooperation. Problems for further investigations are charted

  16. Low Activation Vanadium Alloys for Fusion Power Reactors - the RF Results

    International Nuclear Information System (INIS)

    Full text: The Results of development and researches of functional properties of low activation vanadium alloys (V-Ti-Cr and V-Cr-W-Zr-C systems) being developed for the cores of nuclear fusion and fission (Gen-IV, space) power reactors are presented. Scientific and technological problems of the investigations are related with enhancement of functional properties based on: 1. Special optimized thermal (TT), thermomechanical (TMT) and thermochemical (TCT) treatments of V-4Ti-4Cr alloys. 2. Development of new (V-Cr-W-Zr-C system) vanadium alloys. The TMT and TCT regimes ensuring the capability of significant (up to 2 times) enhancement of yield strength in the temperature range up to 800°C keeping relatively high plasticity reserve have been found for alloys. The results of the theoretical, modeling and simulating studies of characteristics of self-point defects and dislocations, their interactions and mobility are presented. Nuclear physics characteristics (primary radiation damage, activation, transmutation, postreactor cooling) of alloys irradiated for a long time in neutron spectra of the fusion reactor DEMO-RF (15.3 dpa/year) and fast power reactor BN-600 (80 dpa/year) are calculated. The interaction characteristics of V-4Ti-4Cr alloy with hydrogen and the influence of hydrogen on mechanical properties of the alloy (impact toughness, internal friction) have been studied. Obtained results allows one to recommend the vanadium alloys for applications in nuclear reactors at operating temperature window 300 - 800(850)°C. The planes of high-dose and high- temperature reactor tests of vanadium alloys are scheduled at material science assemblies of reactor BN-600 (2013 - 2015, doses 50 - 200 dpa, irradiation temperatures 400 - 800°C). (author)

  17. A neutronic analysis of the dry processed nitride fuel in a sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    Nitride fuel is being considered as an alternative to oxide fuel for fast reactors owing to its high thermal conductivity. Because the nitride fuel is composed of one nitrogen atom per one heavy metal atom instead of two for the oxide fuel, it leads to a hardened neutron spectrum and consequently to a higher breeding ratio. In addition, the nitride fuel has other excellent characteristics such as a high material density (13.5 g/cm{sup 3}), high melting temperature ({approx}2,800 .deg. C), no interaction with a sodiumbond, good compatibility with the cladding material, etc. Recently, researches on the nitride fuel such as the fuel fabrication process and irradiation test have been actively carried out in the U.S., Russia, Japan, etc. In this study, the neutronic feasibility of the dry process nitride fuel cycle is assessed for a sodium-cooled fast reactor (SFR), which was recommended as one of the Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis is performed for two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch (Case 1) and a modified BN-600 core (Case 2). Both cores are composed of two core regions and set as a breakeven core without blankets, which avoids the separation process of transuranic (TRU) elements from the spent fuel and the supply of additional fissile material. In this study, the reactor characteristics such as the TRU enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. are obtained for the equilibrium core under a fixed fuel management scheme.

  18. Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

    International Nuclear Information System (INIS)

    The Federal Target Program (FTP) of Russian Federation 'Nuclear Energy Technologies of the New Generation for 2010-2015 and for Perspective up to 2020' is aimed at development of advanced nuclear energy technologies on the basis of closed fuel cycle with fast reactors. There are advanced fast reactor technologies of the 4. generation with liquid metal cooled reactors. Development stages of maturity of fast sodium cooled reactor technology in Russia includes experimental reactors BR-5/10 (1958-2002) and BOR-60 (since 1969), nuclear power plants (NPPs) with BN-350 (1972-1999), BN-600 (since 1980), BN-800 (under construction), BN-1200 (under development). Further stage of development of fast sodium cooled reactor technology in Russia is commercialization. Lead-bismuth eutectic fast reactor technology has been proven at industrial scale for nuclear submarines in former Soviet Union. Lead based technology is currently under development and need for experimental justification. Current status and prospects of State Corporation 'Rosatom' participation in GIF activities was clarified at the 31. Meeting of Policy Group of the International Forum 'Generation-IV', Moscow, May 12-13, 2011. In June, 2010, 'Rosatom' joined the Sodium Fast Reactor Arrangement as an authorized representative of the Russian Government. It was also announced the intention of 'Rosatom' to sign the Memorandum on Lead Fast Reactor based on Russia's experience with lead-bismuth and lead cooled fast reactors. In accordance with the above FTP some innovative liquid metal cooled reactors of different design are under development in Russia. Gidropress, well known as WER designer, develops innovative lead-bismuth eutectic cooled reactor SVBR-100. NIKIET develops innovative lead cooled reactor BRESTOD-300. Some other nuclear scientific centres are also involved in this activity, e.g. Research and Development Institute for Power Engineering (RDIPE). Optimum

  19. 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)

  20. Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

    Directory of Open Access Journals (Sweden)

    Darnowski Piotr

    2015-03-01

    Full Text Available An analysis of the influence of addition of minor actinides (MA to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-of-life (BOL state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fluxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW decreases with increasing MA inventory in the fuel and confirms that the addition of MA has a negative effect on the delayed neutron fraction.

  1. The fifth research coordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

    International Nuclear Information System (INIS)

    The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The objectives of the fifth RCM were: to review the progress achieved since the 4th RCM; to review and finalize the draft synthesis report on BN-600 MOX Fueled Core Benchmark Analysis (Phase 4); to compare the results of Phase 5 (BFS Benchmark Analysis); to agree on the work scope of Phase 6 (BN-Full MOX Minor Actinide Core Benchmark); to discuss the preparation of the final report. In this context, review and related discussions were made on the following items: summary review of Actions and results since the 4th RCM; finalization of the draft synthesis report on BN-600 full MOX-fueled core benchmark analysis (Phase 4); presentation of individual results for Phase 5 by Member States; preliminary inter-comparison analysis of the results for Phase 5; definition of the benchmark model and work scope to be performed for Phase 6; details of the work scope and future CRP timetable for preparing a final report

  2. 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

  3. 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

  4. Research of 3-D hexagonal nodal transport method for fast reactor

    International Nuclear Information System (INIS)

    The 3-D hexagonal nodal transport theory calculation method for fast reactor core was studied. Based on this method, 3-D hexagonal nodal transport code NAST was developed. The surface average angular fluxes were approximated by an azimuthally symmetric double Pn-expansion DP1 and DP3, and 1-D discrete ordinates equations were solved on a fine spatial mesh within the node. Considering the characteristics of the nodal method, the response matrix method was used in the iterations. Therefore, the calculation within the node was simplified and time was saved. The code was tested for the keff, calculation of CEFR and BN-600. A good agreement with the reference results was achieved. (authors)

  5. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The existing sodium cooled fast reactors (SFR) have two types of designs--loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphenix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL's Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed

  6. 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)

  7. 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.

  8. Modern Approaches to Safety Assurance of a New Generation of Sodium Fast Reactors

    International Nuclear Information System (INIS)

    In the stage of designing sodium cooled fast reactors (SFRs) of a new generation there is a task to improve their inherent safety up to the level higher than that of the previous SFR designs. Modern safety requirements to the SFR of the fourth generation are described. Through the example of the BN-1200 reactor, approaches to safety assurance are demonstrated using development of inherent safety properties up to a brand new level compared to that of the earlier commercial reactor designs (BN-600 and BN-800). Also, passive safety devices and systems applied in the BN-1200 design are described. The goal is to eliminate the necessity for evacuation of residents under conditions of any possible realistic accidents. The paper presents properties of inherent self-protection of the BN-1200 reactor and estimation of their effectiveness in terms of safety assurance. The basic design approaches concerning safety are considered, including additional measures as applied to the BN-1200 reactor. These include measures aimed at the elimination or minimization of sodium leaks; design approaches to the passive shutdown systems (PSS) using various operation principles, namely: hydraulically suspended absorber rods operating in case of coolant flow rate decrease (PSS-H) and absorber rods operating in the case of increase of the core outlet coolant temperature above a certain value (PSS-T); passive decay heat removal system; sodium plenum above the core; gastight compartment above the reactor; core catcher made of refractory metal; and reactor guard vessel. (author)

  9. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  10. Estimation of accuracy for calculation of neutron field distribution in fast reactor on the reactor experiments basis

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V.; Khomyakov, Yu.S.; Kotchetkov, A.L.; Semyonov, M.Yu.; Seryogin, A.S.; Tsyboulya, A.M. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, named after A. Leypunsky (SSC RF - IPPE), 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    Calculation of a neutron field distribution in fast reactor is the basic method for forecasting of such characteristics as local and average energy release values in pins and subassemblies, radiation doze on constructional materials, burning out of fuel. And calculation remains dominating means of the control over these parameters. Accuracy of calculation is numerical expression for reliability of definition of the specified characteristics and base for formation of design limits and reserves. The purpose of the given work consist in estimation of accuracy of modern computing means for calculation by the analysis of the available experimental data received on BFS critical facility (IPPE, Russia) and BN-600 power reactor. The comparison analysis of calculation and experimental data on measurement of fission rates distributions for {sup 235}U, {sup 238}U, {sup 239}Pu at critical assemblies BFS 62-th, BFS 66-th series and integrated quantity of fissions at BN-600 subassemblies by gamma-scanning method has been fulfilled. For carrying out of the analysis the integrated system MODEXSYS has been developed. This system includes up-to-date version of diffusion program TRIGEX, precision program MMKKENO, database of experimental data and benchmark models of experiments. Combination of various calculation methods allowed to reveal methodical uncertainties of calculation. Calculation of energy release distribution in traditional cores of fast reactors with sodium coolant is carried out on the average with accuracy 2-3%, the maximum discrepancies do not exceed 5%. On the border of core and radial blanket the methodical error of diffusion approximation, reaching 10% for {sup 238}U fission reaction rate, is experimentally confirmed. In the traditional uranium blanket an error reach value of 20%, and discrepancies have regular character: calculation data give lower values of reactions rates in relation to experiment. At transition to the steel reflector character of calculation

  11. Evolution of the technical concept of fast reactors. The concept of BREST

    International Nuclear Information System (INIS)

    Having understood that conventional power was limited by available fuel resources, as well as the environmental concern, and willing to use the advantages of defense nuclear power achievements, the development of civil nuclear power was initiated. Scarce supply of uranium has been a matter of concern from the very beginning of nuclear power development, but plutonium produced in the thermal reactors was supposed to be used as fuel for the fast reactors which would not be limited by fuel resources. In order to attain high breeding ratio and high power density, the first generation of fast reactors were designed with sodium coolant, uranium blanket to make up for a decrease in breeding ratio if uranium oxides were used as fuel. Development of nuclear power in the sixties and seventies was followed by stagnation. Lessons learned from a 50-year experience and new conditions set for power industry demand a new concept of fast reactor which would meet a variety of cost-efficiency and safety requirements in their present understanding. Development of fast breeders in Russia began after commissioning of BN-350 and completion of BN-600 design. According to present demands BREST reactors should be designed so as to implement consistently the principles of natural safety without deviation from materials and technology which was proven in defense and civil nuclear power facilities

  12. Development and validation of a fast reactor core burnup code - FARCOB

    Energy Technology Data Exchange (ETDEWEB)

    Mohanakrishnan, P. [Indira Gandhi Centre for the Atomic Research, CDO, Reactor Physics Division, Kalpakkam, TN 603 102 (India)], E-mail: mohana@igcar.gov.in

    2008-02-15

    A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under construction. FARCOB uses centre mesh differencing scheme with triangular meshes in the X-Y plane. Steady state solution results match exactly with those of other reputed codes DIF3D and VENTURE for SNR-300 benchmarks. For burnup simulation, core is divided into radial and axial burnup zones and burnup equations are solved at constant power. Burnable fuel and blanket number densities are found and stored for each mesh, so that the user can alter burnup zones and core geometry after a burnup step. For validation, results of FARCOB has been compared with results of other institutes in two burnup benchmarks (ANL 1000 MWe benchmark and BN-600 hybrid core benchmark). It is found that FARCOB results match well with those of the other institutes.

  13. BN-800 reactor is a new stage in transition to innovative nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Poplavsky, V.M.; Chebeskov, A.N.; Matveev, V.I. [State Scientific Center of the Russian Federation, Institute for Physics and Power Engineering named after A.I. Leypunsky, Obninsk Kaluga region (Russian Federation)

    2007-07-01

    This paper presents the perspectives of nuclear power development in Russia and the reasons why it is necessary to use for that fast reactor technology. Some features of fast reactor technology and main ideas and technical approaches that have been used in the design of the BN-800 sodium cooled fast reactor are given as well. The BN-800 design is based on the BN-600 design with a series of innovative modifications: -) changing the size and structure of the upper axial blanket in order to get a zero or negative sodium void reactivity effect, -) the addition of scram rods based on passive activation, -) the addition of passive system of emergency cooling with sodium-air heat exchangers, -) a special in-vessel catcher envisaged under the core to catch and retain fragments of the core in case of core disruptive accident, and -) an improved earthquake resistance of all the structures. Such issues as possible options of fuel cycle, closing fuel cycle, transuranium element burning, disposal of plutonium being withdrawn from military programs, etc. are discussed as applied to the BN-800 reactor. Some economic considerations in general outline of the BN-800 unit are presented in the paper. It is important to note that the commissioning of the BN-800 reactor was included into the Federal Goal-Oriented Program - Development of nuclear energy-industrial complex of Russia for 2007-2010 and for perspective up to 2015 -, which was approved by the Russian Government in October 2006.

  14. A dry process fuel cycle analysis for a sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Joon Jeong; Gyu, Hong Roh; Hangbok, Choi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2005-07-01

    The Korean nuclear fuel cycle with the dry-processed oxide fuel sodium-cooled fast reactor (SFR) has been studied by the dynamic analysis method. In addition to the SFR, the current operating Pressurized Water Reactor (PWR) and Canadian Deuterium Uranium (Candu) reactors were considered. For the analysis, the equilibrium fuel cycle of the reference core was chosen from a modified BN-600 core. In the reactor scenario, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 1%. In this study, the spent fuel inventory as well as the amount of plutonium, minor actinides (MA) and fission products (FP) of the recycling fuel cycle were estimated and compared to that of the once-through fuel cycle. The results of the once-through fuel cycle have shown that the demand increases up to 64 GWe and total amount of spent fuel is about 102 kt in 2100. When the dry-processed fuel SFR scenario is implemented, it is expected that the total spent fuel inventory can be reduced by about 80%. Also, the plutonium inventory can be reduced by about 25%. However it was found that the SFR scenario does not contribute to the reduction of the MA and FP, which are important when designing a repository. For a further destruction of the MA, an actinide burner can be considered in a future nuclear fuel cycle. (authors)

  15. USSR experience of the safe operation of nuclear power plants with fast reactors

    International Nuclear Information System (INIS)

    Experience of the operation of nuclear power plants with fast reactors in the USSR is based on the results of work with BN-350 and BN-600 reactors. This experience affords evidence of extremely satisfactory safety characteristics from the point of view both of reliable heat removal from the reactor cores and of the hazard to plant personnel, the environment and population at large from exposure to radiation. The paper gives information concerning the power regime at which the facility is operated and about the most characteristic and dangerous situations which have occurred during operation. A comparison is made between a list of the most dangerous initiating events which are analysed in connection with the design of nuclear power stations with fast reactors (in accordance with USSR standards documents now in force) and the events which were observed in the process of operation. Reference is made to the important role of the more probable initiating events in the overall problem of ensuring the safety of nuclear power plants, especially when this is related to action by the staff which is not provided for in instructions and to possible errors on the part of personnel. (author)

  16. The fast reactor with a cold bottom vessel

    International Nuclear Information System (INIS)

    2 kinds of design have been proposed so far for pool-type fast reactors where the vessel contains the full primary circuit with pumps and heat exchangers. First, the vessel suspended to the concrete vault top, a rounded bottom supporting the core (Phenix, Superphenix or the European Fast Reactor Project), and secondly, the half resting vessel, the upper cylindrical wall resting on a ring supported by the basement (BN-600, BN-800 Russian reactors)). A new kind of design is developed in this paper: the cold resting bottom vessel where the flat vessel bottom cooled at less than 90 C. degrees rests on the concrete basement and supports the core dia-grid above a sodium in thermal gradient, by vertical shells and by the vessel wall. This space contains heavy and refractory materials, preventing bottom piercing by corium, hot mix flow of fuel and steel in case of core fusion accident. A concrete vault contains the reactor under ground level. Resting cylindrical vessel provides stability and robustness for core and internal components. A toric collector welded around the dia-grid receives vertical exchangers and pumps, homogenizing temperature. Cooling of structures is not needed. Sodium leaks through the bottom plate are prevented by sodium freezing. Exchangers and pumps are stable, maintained at roof and collector levels. Reduced thermal exchanges provide better efficiency. Arrest exchangers, located around pump columns above the collector, allow a reduced vessel diameter. Simplified structures provide lesser weight and cost. The cold resting bottom may be applied to loop reactors and to lead-cooled reactors

  17. Upper Limits to Americium Concentration in Medium Size Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Y.P.; Wallenius, J. [Royal Institute of Technology (KTH), AlbaNova University Centre, S-106 91, Stockholm (Sweden)

    2009-06-15

    The fastest way to realize transmutation of minor actinides would be using existing reactor types, adding some proper modifications to allow for insertion of MA in the fuel. According to calculations by Fazio and co-workers, the consumption rate of TRU in a low conversion ratio fast reactor may reach 70-75 % of that of an ADS with uranium free fuel [1]. However, americium introduction brings a negative influence on several safety parameters such as {beta}{sub eff}, Doppler coefficient, coolant temperature coefficient and void worth. Therefore the upper limit of americium that can be included into the fuel needs to be carefully evaluated. In this paper, fast reactor fuels with various minor actinide fractions are loaded into a SAS4A model of the semi-commercial BN600 reactor. Unprotected loss of flow (ULOF) and transient over power (UTOP) accidents are modelled using safety parameters obtained from Monte Carlo simulations as well as from the deterministic calculations published by Fazio et al. Applying the latter parameters (obtained with VARI3D), the upper limit to MA concentration in the fuel of a medium sized SFR of BN-600 type appears at 12%, corresponding to 8% of americium. We note however that the Doppler constants displayed by Fazio et al for MA concentrations above 10% have a considerably larger magnitude than those obtained with MCNP. Applying the safety parameters obtained with Monte Carlo simulations and updated nuclear data evaluations, we find that the upper limit to the americium concentration allowing to survive a ULOF is about half of that inferred by the use of parameters from VARI-3D. Since such a difference has a major impact on the predicted americium transmutation capability of SFR, it is of high priority to analyse the reasons for the apparent discrepancies. We note here that the major contribution to the Doppler feedback comes from capture resonance in U-238 and Pu-240 residing below the sodium scattering resonance located at 3 keV, and that

  18. Study of swelling and mechanical properties of 06Kh16N15M2G2TFR steel fuel cans irradiated in NB-600 reactor up to 87.5 dpa dose

    International Nuclear Information System (INIS)

    The diameter change, void swelling and the mechanical properties of the fuel cans fabricated from the 0Kh16N15M2G2TFR cold worked stainless steel and irradiated in the BN-600 reactor up to a peak burnup of 11,5 % h.a. (damage dose of 87,5 dpa) have been studied. The peak diameter change of cans studied is 2,2-3,4 % in the temperature range of 420-460 deg C, the peak ovalization being 0,4 mm. The void swelling rate of the 06Kh16N15M2G2TFR steel in the dose range of 60-87,5 dpa was equal to 0,3 % dpa. With increasing the damage dose up to critical damage level of 87,5 dpa a gradual mechanical property degradation is observed due to the corrosion cracks generation. 5 figs

  19. Chemical operational experience with the water/steam-circuit at KNK II; Presentation at the meeting on Experience exchange on operational experience of fast breeder reactors, Karlsruhe/Bensberg/Kalkar, June 18. - 22. 1990

    International Nuclear Information System (INIS)

    The availability of sodium cooled reactors depends essentially from the safety and reliability of the sodium heated steam generator. The transition from experimental plants with 12-20 MW electrical power to larger plants with 600 MW (BN-600) or 1200 MW (Superphenix) required the change from modular components to larger and compact steam generators with up to 800 MW. Defects of these large components cause extreme losses in availability of the plant and have to be avoided. In view of this request, a comprehensive test program has been performed at KNK II in addition to the normal control of the water/steam-circuit to compile all operational data on the water and steam side of the sodium heated steam generator. This paper describes the plant and the water/steam-circuit with its mode of operation. The experience with the surveillance and different methods of the conditioning are discussed in detail in this presentation

  20. Analyzing the maximum design fault in the core of a fast reactor

    International Nuclear Information System (INIS)

    This paper addresses the problem of design faults for nuclear power stations with fast reactors, mainly the clogging of the cross section of an individual fuel element, caused by swelling of the element itself, by precipitation of foreign substances from the coolant or by penetration by foreign objects. This leads to a reduction in the flow of coolant through the element and damage, destruction or melting of the fuel element with consequential damage to its immediate surroundings. This paper examines some aspects of such a worst-case fault for reactor type BN-600. The development of the fault is observed at the stages of fuel-element overheating and the failure of its seal by observing the delayed neutrons. If the reactor has not already been shut down, the boiling of the sodium and the melting of the fuel can be registered from the neutron-flux and acoustic noise and also by the system for monitoring the reactivity balance, which initiates an alarm signal

  1. A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2006-06-15

    Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted.

  2. A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted

  3. Feedback experience of sodium cooled fast reactors in the world and application to the design of advanced reactors; Bilan de l'experience de fonctionnement des rapides a sodium dans le monde et application a la conception des futurs reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Martin, L. [CEA Valrho, Site de Marcoule (DCP), 30 (France)

    2007-05-15

    18 fast reactors using sodium or sodium-potassium (for 2 reactors) as coolant, operated or operate in the world. They represent a valid feedback experience of 385 cumulated operation years. The analysis of the availability rates shows that these reactors share a common pattern: when they began to operate they faced technological difficulties (due to their prototype role) and then as problems were solved they gradually became more robust. The Russian reactor BN-600 is a good example: it got problems at steam generator level and with fuel but now its availability rate nears 80 per cent. The problems that occurred in fast reactors are: -) water-sodium reactions due to leaks in steam generators, -) the blindly handling under sodium of fuel elements, -) the leaks of sodium in the facility, -) the presence of air and impurities in the primary cooling system, -) cladding failures, -) the use of the austenitic steel 321 and of 15-D3 steel (they were prone to crack). These problems were solved through successive improvements, these improvements represent a precious asset for the design of advanced reactors. (A.C.)

  4. Analysis of Dynamic Regimes at Nuclear Power Plants with Fast Reactors Using the JOKER Code

    Energy Technology Data Exchange (ETDEWEB)

    Seleznev, E.F.; Aizatulin, A.I.; Belov, A.A.; Prianichnikov, A.V.; Fedorov, I.V. [All-Russian Research Institute for NPP Operation (VNIIAES), 25, Ferganskaya str., Moscow, 109507 (Russian Federation); Karpenko, A.I.; Tuchkov, A.M.; Balakhnin, E.V. [Beloyarskaya Nuclear Power Plant (BNPP), BNPP, Zarechnyi, Sverdlovsk region, 624250 (Russian Federation)

    2008-07-01

    To analyze safety of Nuclear Power Plants (NPP) with fast reactors, including studying of different NPP operational modes ranging from normal operation up to hypothetical accidents, a software complex - the JOKER Code [1] - was created simulating the behavior of parameters at BN-type fast reactor under steady-state and transient processes therein through the use of: -reactor core model; and -models of equipment and pipelines of primary, secondary and third circuits of the reactor. The core model contains: a neutron-physical model; a thermal-hydraulics model; and a core thermo-mechanics model. The neutron-physical model is based on the use of spatially distributed kinetics of the core. One-dimensional thermal-hydraulic model with regimes 'before' and 'after' the onset of coolant boiling serves as the thermal-hydraulics model. The thermo-mechanics model includes examination of the behavior of fuel and cladding for the cases of fuel burnup, cracking and melting of both cladding and fuel. The Codes GEFEST (Russia) [2] and SAS-4? (USA) [3] are the JOKER Code's analogues. The GEFEST Code is used at NPP with BN-600 fast reactors to justify safe operation of real fuel loads into reactor installations - mainly for calculations of neutron-physical parameters of the core under steady-state regime; the code has a license of the Russian Supervisory Authority and has been used over many years at Beloyarskaya NPP. The SAS-4? Code developed at ANL (USA) is well known throughout the world as a software complex for analysis of fast reactor projects. (authors)

  5. Reprocessing of research reactor spent nuclear fuel at the PA 'Mayak'

    Energy Technology Data Exchange (ETDEWEB)

    Kolupaev, D

    2007-07-01

    The first Russian reprocessing facility, known as RT-1 (located at PA-Mayak in Ural region) was started on the radiochemical plant base in 1977. Nowadays RT-1 is the sole operating reprocessing plant in Russia. The main features of RT-1 is its broad spectrum of reprocessing spent nuclear fuels (SNF). The following spent fuel types are reprocessed: -) SNF from PWR reactors (WWER-440) and FB reactor (BN-600); -) SNF of transport ship reactors; -) production reactors SNF; -) research reactor spent nuclear fuel. The world-known technological processes are used at RT-1, but there are the following distinctive features. First, the universality of the three technological lines which allows not only the reprocessing of various SNF kinds, but also to implement the combined reprocessing of some types of them. Secondly, extraction of neptunium during SNF reprocessing which is used to implement its separate storage and for radioisotope production. Thirdly, target enrichment of recycled uranium is achieved by mixture of uranium from reprocessing SNF of various kinds. And fourth, the extraction of various elements (such as cesium, strontium, promethium and etc.) which are used for radioisotope production. Concerning research reactor spent fuel reprocessing, the range of used fuels to be processed includes mainly fuel composition on the basis of aluminium and magnesium and can be extended to metal uranium fuels. One of the main issues to overcome is the fact that a large quantity of spent fuel elements from research reactors are leaking because of their long storage in water. This problem rises safety concerns for shipment and for interim storage at the reprocessing plant.

  6. A study on accident prevention of liquid metal reactors through operating experience analysis

    International Nuclear Information System (INIS)

    A demonstration LMR (Liquid Metal Reactor), called as KALIMER (Korea Advanced LIquid MEtal Reactor), has been being developed as part of the nuclear mid and long-term projects of the government since 1997. To ensure the safety of the KALIMER, the capability to cope with accidents must be enhanced by incorporating means and measures to prevent and mitigate accidents into the design of the KALIMER. The means and measures can be found out through analyzing operating experience in LMRs. Therefore, operating experience reported in published literature was collected and analyzed for the following 9 foreign LMRs: MONJU, Superphenix, Phenix, PFR, JOYO, EBR-II, FFTF, BN-350, BN-600. The analyses results show that accidents can be categorized into the following major groups: sodium leakage, sodium fire, sodium-water reaction, abnormal decrease of core reactivity, components vibrations, sodium aerosol deposits. Based on the results of accident cause analysis for each category, the means and measures to prevent and mitigate the each accident category were obtained

  7. PROMILLE database as a part of JNC reactor physics analytical system for BFS-2 fast critical facility experiments analysis

    International Nuclear Information System (INIS)

    The PROMILLE database for experimental data from the BFS-2 fast critical facility (Institute of Physics and Power Engineering (IPPE), Russia) has been developed and embedded into the JNC reactor physics analytical system to provide a strict documentation format, a common data source for different analytical tools and a unique export interface with different reactor codes. PROMILLE should be considered not only as a database but also as a bank of interfaces because of its dynamic role in the analytical process. The database currently accepts data from the simulation materials (pellets, tubes and bars) as well as full cores descriptions. A core description involves all different unit cells forming loading elements, all types of the loading elements forming a layout and the layout itself. In fact it is a description of criticality experiments. Export interfaces for the CITATION-FBR code and the SLAROM and CASUP codes have been developed. The PROMILLE software was developed with MS Visual Basic 6.0 and the data is kept in the data tables generated with the MS Access database management system. Data for eight BFS-2 assembly configurations have been incorporated. They include BFS-58-1i1 uranium-free plutonium assembly with inert material included in its fuel matrix and also seven BFS-62 modifications simulating different stages of investigation of MOX fuel based BN-600 core. (author)

  8. Nuclear power based on fast reactors. Scientific idea, early experience, new start

    International Nuclear Information System (INIS)

    Full text: In the early 1950s the author joined the IPPE activities on the theory of reactors (First World NPP, nuclear submarines). In the late 1950s A.I. Leypunsky placed him in charge of the research for the NPP development effort. He was also the research supervisor for the BN-350 and BN-600 fast reactor projects (in the 1960s) and for the BN-350 reactor start (in the 1970s). Since the accidents at EBR-I and the E.Fermi NPP, these became the first successful embodiments of the Fermi idea (1944), that is, nuclear power based on fast reactors. The BN-350 reactor operated for 25 years and the BN-600 is still in operation. However, none of the projects have been continued. As the result of comprehending (already at the Kurchatov Institute) the causes for the unsuccessful early experience, in the 1980s the author gave up the Fermi-originated concept of the started on the Pu from thermal reactors and embarked on the development (at NIKIET) of the BREST of a moderate power rating with BR∼1 to operate on enriched U or Pu. The consumption of U (and the separation work) to start the FR on enriched U is considerably below that for thermal reactors-generated Pu, and the FR natural safety properties with respect to accidents, wastes and proliferation resistance once the adequate technology is selected (nitride fuel of an equilibrium composition, on-site processing of fuel, Pb in place of Na) also make large NPPs much cheaper. High rates of Pu breeding are therefore unnecessary, while U is used in full with BR ∼1, that is, 100-200 times as effectively as in thermal reactors, so inexhaustible low-grade ores suit as well. Fitting FRs with a Th-blanket in future will also provide Th-3U fuel for FRs of small-sized NPPs for local needs. Still, the prime task of nuclear power will remain generation of electricity at large NPPs, where it is profitable to use FRs in closed fuel cycles. The growth of nuclear power will entail an increase in the share of electricity in total

  9. Evolution of magnetic properties of cladding austenitic steel under irradiation in a reactor

    Science.gov (United States)

    Chukalkin, Yu. G.; Kozlov, A. V.; Evseev, M. V.

    2014-03-01

    Magnetic properties of samples of austenitic steel ChS-68 cut from the cladding of a fuel element, which was irradiated in a BN-600 fast-neutron reactor to a maximal damage dose of ˜80 displacements per atom (dpa) at temperatures of 370-587°C, have been investigated. It has been established that irradiation with fast neutrons leads to the formation of ferromagnetic microregions, the effective sizes and concentration of which depend on the damage dose. It has been shown that, at damage doses higher than ˜55 dpa, small spontaneous magnetization and magnetization hysteresis, which are characteristic of the ferromagnetic state, appear in the samples. It is assumed that the ferromagnetic microregions are the nuclei of the α' phase and the radiation-induced segregation microregions, in which the spacing between the nearest iron atoms exceeds the critical distance that determines the change in the sign of exchange interaction. Arguments in favor of this assumption are presented.

  10. The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mitenkova, E. F.; Novikov, N. V. [Nuclear Safety Inst. of Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Blokhin, A. I. [State Scientific Center of Russian Federation, Inst. of Physics and Power Engineering Named after A.I. Leypunsky, Bondarenko Square 1, Obninsk, Kaluga Region, 249030 (Russian Federation)

    2012-07-01

    The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

  11. The state of the art of designing complex-alloyed steel EhK164 for BN type reactor fuel cans

    International Nuclear Information System (INIS)

    Aimed to achieve higher fuel burnups in fast reactors a complex-alloyed with titanium, niobium and vanadium steel EhK164 (07Kh16N19M2G2BTFR) is designed as a material for fuel cans. Steel EhK164 is developed on the basis of steel 172-U exhibiting a good radiation resistance up to burnups of 10% h.a. The tests show that steel EhK164 possesses greater structure and phase stability on ageing and under irradiation. Short-term mechanical properties in the temperature range of 20-700 deg C and creep characteristics of cold rolled cladding tubes at temperatures of 600-700 deg C are rather close for both steels. Simulation experiments on steel irradiation with chromium ions up to 140 dpa damaging dose show that steel EhK164 has a better resistance to swelling compared to steel 172-U. It is concluded that a newly designed steel EhK164 in a cold worked state can be recommended as material for fuel cans in experimental fuel assemblies of BN-600 reactor with the aim of attaining fuel burnup ≥ 12% h.a. and damaging dose ≥ 100 dpa

  12. 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

  13. Technical meeting on 'Operational and decommissioning experience with fast reactors'. Working material

    International Nuclear Information System (INIS)

    For three decades, several countries had large and vigorous fast breeder reactor development programs. In most cases, fast reactor development programs were at their peaks by 1980. Fast test reactors [Rapsodie (France), KNK-II (Germany), FBTR (India), JOYO (Japan), DFR (UK), BR-10, BOR-60 (Russia), EBR-II, Fermi, FFTF (U.S.A.)] were operating in several countries, with commercial size prototype reactors [Phenix, Superphenix (France), SNR-300 (Germany), MONJU (Japan), PFR (UK), BN-350 (Kazakhstan), BN-600 (Russia)] just under construction or coming on line. From that time onward, fast reactor development in general began to decline. By 1994 in the USA, the Clinch River Breeder Reactor (CRBR) had been cancelled, and the two fast reactor test facilities, FFTF and EBR-II had been shutdown - with EBR-II permanently, and FFTF in a standby condition. Thus, effort essentially disappeared for fast breeder reactor development. Similarly, programs in other nations were terminated or substantially reduced. In France, Superphenix was shut down at the end of 1998; SNR-300 in Germany was completed but not taken into operation, and KNK-II was permanently shut down in 1991 after 17 years of operation, and is scheduled to be dismantled by 2004; in the UK, PFR was shut down in 1994; BN-350 in Kazakhstan was shut down in 1998. It is difficult to argue that fast breeder reactors will be built in the near term when no commercial market exists and there is a plentiful supply of cheap uranium. Nevertheless, it is reasonable to assume that, were nuclear energy to remain an option as part of the long-term world energy supply mix, meeting the sustainability requirements vis-a-vis natural resources and long-lived radioactive waste management will require deploying systems involving several reactor types and fuel cycles operating in symbiosis. Apart from cost effectiveness, simplification, and safety considerations, a basic requirement to these reactor types and fuel cycles will be flexibility

  14. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, EC-JRC, Westerduinweg 3, P.O. Box 2, NL-0 1755 ZG Petten (Netherlands)

    2006-07-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  15. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  16. Research co-ordination meeting on updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactor reactivity effects. Working material

    International Nuclear Information System (INIS)

    The main purpose of the second RCM of the CRP on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects' was to present and discuss the results of Phases 1 and 2 of the CRP and to agree on the work plan for Phase 3. The RCM considered: The presentation of work performed under Phase 1 and 2 of the CRP by CIAE (China), AEAT/CEA joint European contribution), IGCAR (India), JNC (Japan), KAERI (Republic of Korea), OKBM and IPPE (Russia), and ANL (USA); A preliminary analysis of the results from Phases 1 and 2; The definition of the work to be performed under Phase 3 of the CRP; A discussion of the Phase 4 CRP work scope. Calculation results of the proposed benchmark for a hybrid UOX/MOX fuelled core of the BN-600 reactor are resented. The calculation results of all the participants include effective multiplication factors obtained by both diffusion and Monte Carlo methods; fuel Doppler constants; steel Doppler constants; sodium density coefficient; steel density coefficients; fuel density coefficient; absorber density coefficient; axial and radial expansion coefficients; dynamic parameters; power distribution

  17. Operational and decommissioning experience with fast reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    aspect of this knowledge base is given by the accumulated operational experience. The participants in the 33rd and the 34th Annual Meeting of the International Working Group on Fast Reactors, recommended holding a technical meeting (TM) on Feedback from Operational and Decommissioning Experience with Fast Reactors, and to launch a Co-ordinated Research Project (CRP) on Generalization and Analyses of Operational Experience with Fast Reactor Equipment and Systems (Preserve Fast Reactor Operation and Decommissioning Experience). The scope of the TM was to provide a global forum for information exchange on fast reactor operational and decommissioning experience. The objectives of the TM were to: Exchange detailed technical information on fast reactor operation and/or decommissioning experience with DFR, PFR (UK); KNK-II (Germany); Rapsodie, Phenix, Superphenix (France); BR-10, BOR-60, BN-600 (Russian Federation); BN-350 (Kazakhstan); SEFOR, EBR-II, Fermi, FFTF (USA.); FTBR (India); JOYO, MONJU (Japan); Present the status of the work concerning the knowledge preservation efforts related to the experience accumulated in the various member states from the operation and decommissioning of fast reactors; Start the preparation of the planned Co-ordinated Research Project (CRP) on Generalization and Analyses of Operational Experience with Fast Reactor Equipment and Systems

  18. A problem to determine short term mechanical properties changes of ferrite-martensite and austenitic steels as materials of fuel assembly of fast reactors under high dose neutron irradiation

    International Nuclear Information System (INIS)

    The results of mechanical tests of flat and ring-shaped samples of two ferrite-martensite steels C0.1-Cr13-Mo2-Nb-V-W and C0.1-Cr12-Mo-Nb-V-W irradiated to different damage doses (up to 100 dpa) have been performed in this work. It have been shown that values of plasticity and strength characteristics determined on this sample types are different. Specific elongation takes the values 8-12% for the flat samples, at the same time, t takes the values 1-3% for the ring-shaped samples at room temperature. A character of fluence dependence of mechanical properties is identical. The steels show viscous damage in all tests. Samples of fuel pin cladding fabricated from the austenitic steel C0.1-Cr16-Ni15-Mo3 were also investigated after there working out in BN-600 reactor up to 76 dpa. Ring-shaped samples were tested at standard single-axle tension. Tube samples were tested by internal pressure of solid filler. All samples were fabricated from one and the same section, mechanical properties obtained are different. Specific elongation of the brittlest section of fuel pin was 0-0.9% for the ring-shaped samples and 2-7% for the tube samples at room temperature. Fractographic investigations were carried out on the samples after mechanical tests. Possible reasons of such difference have been discussed in the work. (author)

  19. N Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  20. The fourth research co-ordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

    International Nuclear Information System (INIS)

    The fourth Research Co-ordination Meeting (RCM) of the Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effect' was held during 19-23 May, 2003 in Obninsk, Russian Federation. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The first RCM took place in Vienna on 24 - 26 November 1999. The meeting was attended by 19 participants from 7 Member States and one from an international organization (France, Germany, India, Japan, Rep. of Korea, Russian Federation, the United Kingdom, and IAEA). The participants from two Member States (China and the U.S.A.) provided their results and presentation materials even though being absent at the meeting. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN- 600 core were evaluated. Contributions of the participants in the benchmark analyses is shown. This report first addresses the benchmark definitions and specifications given for each Phase and briefly introduces the basic data, computer codes, and methodologies applied to the benchmark analyses by various participants. Then, the results obtained by the participants in terms of calculational uncertainty and their effect on the core transient behavior are intercompared. Finally it addresses some conclusions drawn in the benchmarks

  1. Reactor Physics

    International Nuclear Information System (INIS)

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  2. Reactor Physics

    International Nuclear Information System (INIS)

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  3. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  4. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  5. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  6. Estimation of the effect of operation on deformation of BN-600 fuel assembly tubes and fuel pin cladding fabricated of austenitic Cr/Ni steel

    International Nuclear Information System (INIS)

    The formulae given in this paper describe relationship between structural elements deformations and operational factors (i.e. temperature, damage doze). Measurement results obtained following irradiation of fuel cladding (type Kh16N15M3B steel, austenitized) and fuel assembly tubes (type O8Kh16N11M3 mechanically and thermally treated) are in good agreement with the design values. Based on these formulae models to assess temperature of fuel claddings and fuel assembly tubes during irradiation have been developed. In particular it is proposed to use the radiation of elongated cladding to maximum increased of its diameter. These models allow to assess temperature deviations in the core and inside separate fuel assemblies. 4 refs.; 6 figs. (author)

  7. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  8. Research reactors

    International Nuclear Information System (INIS)

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  9. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  10. Research reactors

    International Nuclear Information System (INIS)

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted

  11. Reactor container

    International Nuclear Information System (INIS)

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  12. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  13. Heterogeneous reactors

    International Nuclear Information System (INIS)

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author)

  14. Status and perspectives of fuel developments for fast neutron reactors of 4th generation

    International Nuclear Information System (INIS)

    challenger to oxide fuel. In such a context, the merits and drawbacks of the metal fuel option for large SFR cores must be re-assessed, and its performances compared with that of oxide and carbide/nitride fuels. This paper summarizes the current status of fuel development and perspectives. Basic features of oxide, metal and other fast reactor fuels (carbide and nitride) are compared from the viewpoints of fuel cycle (fresh fuel fabrication and spent fuel treatment), in-pile behaviour, core performances, and safety. The paper also briefly reviews the potential offered by innovative structural materials developed for high temperature resistance (SiC, refractory metals) for the GFR, or low swelling behaviour under irradiation (ODS,...) for the SFR. The role of experimental reactors is underlined for further assessment of the in-pile behaviour of fuels with representative materials and realistic conditions (burn-up, MA content, neutron flux...). An optimal use of existing irradiation reactors (Phenix, Joyo, Monju, BOR-60, BN-600) is necessary until new reactors, under construction (JHR, CEFR, PFBR) or planned (ALLEGRO, ASTRID) can be put in operation. The paper pleads for the implementation of multilateral collaboration at the European and broader international levels for a continuous capability of innovative fuel qualification. (author)

  15. Plasma reactor

    OpenAIRE

    Molina Mansilla, Ricardo; Erra Serrabasa, Pilar; Bertrán Serra, Enric

    2008-01-01

    [EN] A plasma reactor that can operate in a wide pressure range, from vacuum and low pressures to atmospheric pressure and higher pressures. The plasma reactor is also able to regulate other important settings and can be used for processing a wide range of different samples, such as relatively large samples or samples with rough surfaces.

  16. Reactor physics

    International Nuclear Information System (INIS)

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  17. Compact Reactor

    International Nuclear Information System (INIS)

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  18. Computational software package for analyzing the fast neutron reactor safety: Its improvement and development prospects

    International Nuclear Information System (INIS)

    Full text: Currently, ROSATOM institutions (and RF SSC IPPE first of all) are developing a software package executing major fast sodium-cooled neutron reactor safety analysis. This study field includes subroutines which permit to carry out 3-D calculation runs for inter-related unsteady neutronics, thermohydraulics, thermomechanical in-reactor and whole-facility accident processes involving sodium boiling, in-core steel and fuel meltdown and relocation with account of multi-component multi-velocity thermally-unsteady state models, in-core corium containment, reactor pressure vessel and in-core structure strength under emergency conditions, software for computing sodium fire and in-SG sodium-water interaction consequences, on-site radioactive contamination propagation and estimating the feasible NPP near-site population irradiation. The paper does not discuss the software purposed for analyzing technological aspects of fast neutron reactor safety (e.g. sodium combustion, sodium-water interaction in a steam generator) and software for emergency state reactor neutronics parameters. The paper issue is an integral software package needed to carry out whole in-core and NPP thermohydraulic processes computation. Named computational codes were developed after large experimental data array which have been acquired during several decades at the domestic and foreign reactors and test benches and thereafter accumulated from public domain publications available now. These experimental data constitute the analysis base; they have been implemented to create so-called flow pattern diagrams, closing relations of the computational codes. These computational codes have been proposed by the RF SSC IPPE and have been successfully implemented when reviewing and processing experiment run data acquired in the course of the emergency/transition state studies at BR-10, BN-350, BN-600 reactors, domestic facilities, and after analysis of some dedicated foreign reactor and facilities (e

  19. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  20. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

  2. BOR-60 reactor as an instrument for experimental substantiation of fuel rods for advanced NPPs

    International Nuclear Information System (INIS)

    . The principle task was to provide the required temperature conditions on specimens. This was achieved through the use of the thermal insulation gaps, intense cooling or additional heating at the expense of radiation energy release or fuel fission. As a result the reactor is used for testing various advanced types of fuel and structural materials at high thermal loads (100kW/m), temperatures (100 deg. C), burnups (33% h.a.) and fluences (1.8·1023cm-2 with E>0.1MeV). In case of necessity, the temperature can be stabilized by changing the thermal resistance in the heat transfer or heat removal intensification scheme using the liquid metal kept in the boiling condition. These units also provide the specified height and azimuthal temperature nonuniformity. The experimental facilities can be used for testing the fuel rods of up to 15 mm in diameter placed in different grids (triangular, square etc.) and environments (sodium, lithium, lead, various gases etc.). Due to the availability of the experimental facility for reprocessing of irradiated fuel and production of fast reactor fuel and fuel rods, the reactor is used for the experiments related the closed fuel cycle, such as testing of refabricated fuel with involvement of minor actinides and long-lived fission products into the fuel cycle. The BOR-60 demonstrated an effective operation as a MA burner as well as a power and weapon grade burner. This allows us to solve the important tasks of the nuclear power engineering, in particular to reduce the fuel cost and the quantity of radioactive waste and improve the environmental situation. The BOR-60 reactor has great experience in irradiation of oxide, metal, ceramic, carbide and nitride fuel compositions for reactors of different purposes, in particular for fast sodium reactors. Such fuel properties as regularities of gas release, shape change and structure formation were studied. The results obtained allowed us to substantiate the use of fuel rods for the BN-350 and BN-600

  3. Multifunctional reactors

    OpenAIRE

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much emphasis in research in the last decade. A survey is given of modern developments and the first successful applications on a large scale. It is explained why their application in many instances is ...

  4. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  6. Breeder reactors

    International Nuclear Information System (INIS)

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

  7. Ideas in support to the definition of the Phase 6

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fuelled core of the BN-600 reactor was endorsed as an international benchmark. Phases 1 and 2 consist of RZ and HEX-Z homogeneous models of the hybrid version of the BN-600 reactor. Phase 3 consists of RZ and HEX-Z heterogeneous models of the hybrid version of the BN-600 reactor. Phase 4 consists of RZ and HEX-Z heterogeneous models of the full MOX version of the BN-600 reactor. Phase 5 consists of the Analysis of BFS-62 hybrid configuration in support to Phase 3 studies. The background strategy was defined to make the world safer by using weapon grade Plutonium for civil application. Make that use safe by checking the behaviour of the BN-600 core with limited (hybrid core: Phases 1, 2 and 3) and then full use of MOX (Phase 4); Verify uncertainties on reactivity coefficients and especially on SVRE with some BFS-62 experiments (Phase 5) and use of Minor Actinides in the fuel (Phase 6 and possibly Phase 7). The French Strategy was make the link between existing reactors PWR and GEN-IV ones. From 2030 - 2040, Introduction of 4th generation systems was planned. The P4 and N4 PWR reactors will reach 40 years lifetime at 2025-2035. Lifetime extension to 50 years is considered. The replacement of PWR reactors by Gen IV systems will be effective. Proposal of Phase 6 considers to develop a strategy in connection with GEN IV criteria, use BN-600 as a demonstrator of GEN IV cores, use spent fuels from WWERs, RBMKs as a fuel for use in LMFBR (BN-600 being the first in the row). In Russia, there are roughly 9 GWe WWER and 10.2 GWe RBMK reactors. UOX is being used (no MOX being used), burn up rate is 45 GWd/ton. At the moment, no reprocessing is performed but a reasonable scenario is to develop a simplified dry reprocessing or a dry reprocessing to extract both MA and Pu resulting in no separation and limited Proliferation. Pu vector will no longer be weapon grade. There will be no blanket as far as possible. Study the BN-600 behaviour with this type of fuel

  8. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  9. Reactor utilization

    International Nuclear Information System (INIS)

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  10. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

  14. Program of quality management when fabricating fast reactor vibropack oxide fuel pins

    International Nuclear Information System (INIS)

    There are presented main principles of creation and operation of Quality Management Program in fabricating vibropack oxide fuel pins for BOR-60 and BN-600 being in force in SSC RF RIAR. There is given structure of documentation for QS principal elements. Under Quality System there are defined all the procedures, assuring that fuel pin meets the normative requirements. The system model is complied with the standard model IS 9001. There are shown technologic flowchart and check operation, statistic results of pin critical parameter check as well as main results of in-pile tests. (author)

  15. Reactor container

    International Nuclear Information System (INIS)

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  16. Reactor building

    International Nuclear Information System (INIS)

    The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  18. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  1. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  2. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  4. Nuclear research reactors

    International Nuclear Information System (INIS)

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.)

  5. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

  7. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  8. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  9. Reactor container

    International Nuclear Information System (INIS)

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  10. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  11. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  12. RB reactor noise analysis

    International Nuclear Information System (INIS)

    Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed

  13. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  14. Reactor Physics Training

    International Nuclear Information System (INIS)

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  15. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

  16. Safeguarding research reactors

    International Nuclear Information System (INIS)

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  17. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  18. Nuclear reactor building

    International Nuclear Information System (INIS)

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  19. Reactor Physics Programme

    International Nuclear Information System (INIS)

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  20. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  1. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  2. Process heat reactors

    International Nuclear Information System (INIS)

    The consumption of heat, for industrial and domestic needs, takes up half of the national energy supply; direct utilization of the heat produced by nuclear reactors could therefore contribute to reduce the deficit in the energetic results. The restraints proper to heat consumption (dispersal and variety of consumers, irregular demand) involve the development of the heat transport system structures and adequate nuclear reactors. With this in view, the Commissariat a l'Energie Atomique and Technicatome are developing the CAS reactor series, pressurized water reactors (PWR), (CAS 3G reactor with a power of 420 MW.th.), and the Thermos reactor (100 MW.th.), directly conceived to produce heat at 1200C and whose technology derives from the experimental pool reactors type. In order to prove the value of the Thermos design, an experimental reactor should soon be constructed in the Saclay nuclear research centre

  3. Reactor System Design

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. The Chernobylsk reactor accident

    International Nuclear Information System (INIS)

    The construction, the safety philosophy, the major reactor physical parameters of RBMK-1000 type reactor units and the detailed description of the Chernobylsk-4 reactor accident, its causes and conclusions, the efforts to reduce the consequences on the reactor site and in the surroundings are discussed based on different types of Soviet documents including the report presented to the IAEA by the Soviet Atomic Energy Agency in August 1986. (V.N.)

  6. Zero energy reactor 'RB'

    International Nuclear Information System (INIS)

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  7. High solids fermentation reactor

    Science.gov (United States)

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  8. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  9. Fusion reactor studies

    International Nuclear Information System (INIS)

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  10. Reactor power measuring device

    International Nuclear Information System (INIS)

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  11. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

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

  12. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  13. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  14. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  15. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPRTM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  16. The Maple reactor project

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  17. High temperature reactors

    International Nuclear Information System (INIS)

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements

  18. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  19. Updated codes and methods to reduce the calculational uncertainties of the LMFR reactivity effects. First research co-ordination meeting. Working material

    International Nuclear Information System (INIS)

    The main purpose of the research co-ordination meeting (RCM) was to discuss and agree upon the specification of a benchmark proposal prepared by the Russian delegation (joint IPPE and OBMK) effort. The objective of the CRP was to validate, verify and improve methodologies and computer codes used for calculation of reactivity coefficients in fast reactors aiming and enhancing the utilization of plutonium and minor actinides. The International Working Group on Fast Reactors endorsed the hybrid BN-600 reactor core as a calculational model, thus clearly emphasising the relevance of the CRP in view of the utilization of weapon grade plutonium for energy production in a mixed UOX/MOX core of the BN-600 reactor. Considering the general objective it was agreed that specific objective of the CRP was the comparison whenever possible backed by experimental results, of calculated LMFBR reactivity coefficients and their effect on core transient behaviour

  20. Reactor Safety: Introduction

    International Nuclear Information System (INIS)

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

  1. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  2. Thai research reactor

    International Nuclear Information System (INIS)

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  3. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  4. Tenth annual meeting, Vienna, Austria, 29 March - 1 April 1977. Summary report. Part III

    International Nuclear Information System (INIS)

    The Summary Report - Part III of the Tenth Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercial development of FBRs according to national plans, mostly related to technology problems of containment design, fuel fabrication, fuel failures, sodium pressure, fuel-sodium interaction, computer codes needed for licensing. Most of the discussions were related to the existing reactors: BN-600, BN-350, BN-1600, BOR-60, RAPSODIE, PHENIX

  5. Transmutation of Americium in Fast Neutron Facilities

    OpenAIRE

    Zhang, Youpeng

    2011-01-01

    In this thesis, the feasibility to use a medium sized sodium cooled fast reactor fully loaded with MOX fuel for efficient transmutation of americium is investigated by simulating the safety performance of a BN600-type fast reactor loaded with different fractions of americium in the fuel, using the safety parameters obtained with the SERPENT Monte Carlo code. The focus is on americium mainly due to its long-term contribution to the radiotoxicity of spent nuclear fuel and its deterioration on c...

  6. Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part III

    International Nuclear Information System (INIS)

    The Summary Report - Part III of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercialization LMFBRs according to national plans, mostly related to technology of fuel fabrication, PHENIX fuel pins testing, heterogeneous cores, in service inspection of fuel elements, regulations and licensing, and related OECD activities. Most of the discussions were related to the existing reactors: BR-10, BN-600, BN-350, BN-1600, RAPSODIE and PHENIX

  7. Fast breeder nuclear power plants: outline of nuclear energy development

    International Nuclear Information System (INIS)

    This article describes briefly the route travelled by the Soviet scientists from the time of the principal basic technological options up to the recent period where the reactor BN 600 has been commissioned. The principal changes which will distinguish the next generation, the reactor BN 800 are indicated and the importance, from the point of view of the accompanying economic growth, of obtaining in good time higher breeding ratios is emphasized

  8. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

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

  9. Back-to-back technical meetings (TMs): 'TM on the coordinated project (CRP) analyses of and lessons learned from the operational experience with fast reactor equipment and systems' and 'TM to coordinate the Agency's fast reactor knowledge preservation international project in Russia'. Working material

    International Nuclear Information System (INIS)

    Since the early 1960's, several countries have undertaken important fast breeder reactor development programs. Fast test reactors were constructed and successfully operated in a number of countries, including Rapsodie (France), KNK-II (Germany), FBTR (India), JOYO (Japan), DFR (UK), BR-10, BOR-60 (Russia), and EBR-II, Fermi, FFTF (USA). This was followed by commercial size prototypes (Phenix, Superphenix (France), SNR-300 (Germany), MONJU (Japan), PFR (UK), BN-350 (Kazakhstan), BN-600 (Russia)], either just under construction, coming on line, or experiencing long term operation. However, from the 1980s onward, and mostly for economical and political reasons, fast reactor development in general began to decline. By 1994, in the USA, the Clinch River Breeder Reactor (CRBR) had been cancelled, and the two fast reactor test facilities, FFTF and EBR-II had been shutdown - EBRII permanently, and FFTF, until recently, in standby condition, but now also facing permanent closure. Thus, in the U.S., effort essentially disappeared for fast breeder reactor development. Similarly, programs in other nations were terminated or substantially reduced. In France, Superphenix was shut down at the end of 1998; SNR-300 in Germany was completed but not taken into operation, and KNK-II was permanently shut down in 1991 (after 17 years of operation) and is scheduled to be dismantled by 2004. In the UK, PFR was shut down in 1994, and in Kazakhstan, BN-350 was shut down in 1998. As the interest and activity in the fast breeder reactor diminished, the retirement of many of the developers and acknowledged experts of this technology reached its peak, between 1990 and 2000. The effort and investment required to replace these skills also diminished in parallel. In addition, the facilities (e.g., hot cells, fuel fabrication and inspection lines, seismic test rigs) required to develop and maintain the fast reactor program are drifting into a degraded state or are being shut down. This leads to the

  10. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  11. Evaluation of research reactors

    International Nuclear Information System (INIS)

    The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially

  12. Multipurpose research reactors

    International Nuclear Information System (INIS)

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  13. The nuclear soliton reactor

    International Nuclear Information System (INIS)

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  14. 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...

  15. Fusion reactor research

    International Nuclear Information System (INIS)

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  16. 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

  17. The replacement research reactor

    International Nuclear Information System (INIS)

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  18. PFBR reactor protection

    International Nuclear Information System (INIS)

    Design philosophy adopted for Prototype Fast breeder Reactor (PFBR) is a classical one and has the following features: triplicated sensors for measuring important safety parameters; two independent reactor protection Logic Systems based on solid state devices; reactivity control achieved by control rods; gas equipped modules at the core blanket interface providing negative reactivity. Design verification of these features showed that safety of the reactor can be achieved by a traditional approach since the inherent features of LMFBR make this easy

  19. Reactor BR2

    International Nuclear Information System (INIS)

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  20. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  1. Reactor Safety Analysis

    International Nuclear Information System (INIS)

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

  2. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  3. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  4. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  5. New reactor concepts

    International Nuclear Information System (INIS)

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  6. Reactor construction steels

    International Nuclear Information System (INIS)

    The basic functions of light water reactor components are shown on the example of a pressurized water reactor and the requirements resulting therefrom for steel, the basic structural material, are derived. A detailed analysis of three main groups of reactor steels is presented and the applications are indicated of low-alloyed steels, high-alloyed austenitic steels, and steels with a high content of Ni and of alloying additions for steam generator pipes. An outline is given of prospective fast breeder reactor steels. (J.K.)

  7. Mirror fusion reactors

    International Nuclear Information System (INIS)

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

  8. Natural convection type reactor

    International Nuclear Information System (INIS)

    In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)

  9. INVAP's Research Reactor Designs

    International Nuclear Information System (INIS)

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors IAEA safety

  10. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention provides a control device which can conduct scram and avoid lowering of the power of a nuclear power plant upon occurrence of earthquakes. Namely, the device of the present invention comprises, in addition to an existent power control device, (1) an earthquake detector for detecting occurrence and annihilation of earthquakes and (2) a reactor control device for outputting control rod operation signals and reactor core flow rate control signals depending on the earthquake detection signals from the detector, and reactor and plant information. With such a constitution, although the reactor is vibrated by earthquakes, the detector detects slight oscillations of the reactor by initial fine vibration waves as premonitory symptoms of serious earthquakes. The earthquake occurrence signals are outputted to the reactor control device. The reactor control device, receiving the signals, changes the position of control rods by way of control rod driving mechanisms to make the axial power distribution in the reactor core to a top peak type. As a result, even if the void amount in the reactor core is reduced by the subsequent actual earthquakes, since the void amount is moved, effects on the increase of neutron fluxes by the actual earthquakes is small. (I.S.)

  11. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  12. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  13. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  14. One piece reactor removal

    International Nuclear Information System (INIS)

    Japan Research Reactor No.3 (JRR-3) was the first reactor consisting of 'Japanese-made' components alone except for fuel and heavy water. After reaching its initial critical state in September 1962, JRR-3 had been in operation for 21 years until March 1983. It was decided that the reactor be removed en-bloc in view of the work schedule, cost and management of the reactor following the removal. In the special method developed jointly by the Japanese Atomic Energy Research Institute and Shimizu Construction Co., Ltd., the reactor main unit was cut off from the building by continuous core boring, with its major components bound in the block with biological shield material (heavy concrete), and then conveyed and stored in a large waste store building constructed near the reactor building. Major work processes described in this report include the cutting off, lifting, horizontal conveyance and lowering of the reactor main unit. The removal of the JRR-3 reactor main unit was successfully carried out safely and quickly by the en-block removal method with radiation exposure dose of the workers being kept at a minimum. Thus the high performance of the en-bloc removal method was demonstrated and, in addition, valuable knowhow and other data were obtained from the work. (Nogami, K.)

  15. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  16. The fusion reactor

    International Nuclear Information System (INIS)

    Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

  17. Polymerization Reactor Engineering.

    Science.gov (United States)

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  18. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  19. Gas-cooled reactors

    International Nuclear Information System (INIS)

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  20. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  1. Reactor Safety Analysis

    International Nuclear Information System (INIS)

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

  2. Light water type reactor

    International Nuclear Information System (INIS)

    The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)

  3. Naval propulsion reactors

    International Nuclear Information System (INIS)

    This article deals with the design and exploitation of naval propulsion reactors, mainly of PWR-type. The other existing or conceivable types of reactors are also presented: 1 - specificities of nuclear propulsion (integration in the ship, marine environment, maneuverability, instantaneous availability, conditions of exploitation-isolation, nuclear safety, safety authority); 2 - PWR-type reactor (stable operation, mastered technology, general design, radiation protection); 3 - other reactor types; 4 - compact or integrated loops architecture; 5 - radiation protection; 6 - reactor core; 7 - reactivity control (core lifetime, control means and mechanisms); 8 - core cooling (natural circulation, forced circulation, primary flow-rate program); 9 - primary loop; 10 - pressurizer; 11 - steam generators and water-steam secondary loop; 12 - auxiliary and safety loops; 13 - control instrumentation; 14 - operation; 15 - nuclear wastes and dismantling. (J.S.)

  4. Iris reactor conceptual design

    International Nuclear Information System (INIS)

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  5. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 1014 n/cm2/sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  6. Reactor core monitoring method

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Michitsugu [Tokyo Electric Power Co., Inc. (Japan); Kanemoto, Shigeru; Enomoto, Mitsuhiro; Ebata, Shigeo

    1998-05-06

    The present invention provides a method of monitoring the state of coolant flow in a reactor of a BWR power plant. Namely, a plurality of local power region monitors (LPRM) are disposed to the inside of the reactor core for monitoring a power distribution. Signals of at least two optional LPRM detectors situated at positions different in axial or radial positions of the reactor core are obtained. General fluctuation components which nuclear hydrothermally fluctuate in overall reactor core are removed from the components of the signals. Then, correlational functions between these signals are determined. The state of coolant flow in the reactor is monitored based on the correlational function. When the axial flowing rate and radial flow interference are monitored, the accuracy upon monitoring axial and radial local behaviors of coolants can be improved by thus previously removing the general fluctuation components from signals of LPRM detectors and extracting local void information near to LPRM detectors at high accuracy. (I.S.)

  7. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  8. Mirror reactor surface study

    International Nuclear Information System (INIS)

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  9. FBR type reactor

    International Nuclear Information System (INIS)

    A circular neutron reflector is disposed vertically movably so as to surround the outer circumference of a reactor core barrel. A reflector driving device comprises a driving device main body attracted to the outer wall surface of the reactor barrel by electromagnetic attraction force and an inertia body disposed above the driving device main body vertically movably. A reflector is connected below the reactor driving device. At the initial stage, a spontaneous large current is supplied to upper electromagnetic repulsion coils of the reflector driving device, impact electromagnetic repulsion force is caused between the inertia body and the reflector driving device, so that the driving device main body moves downwardly by a predetermined distance and stopped. The reflector driving device can be lowered in a step-like manner to an appropriate position suitable to restart the reactor during stoppage of the reactor core by conducting spontaneous supply of current repeatedly to the upper electromagnetic repulsion coils. (I.N.)

  10. TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  11. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  12. Nuclear reactor design

    CERN Document Server

    2014-01-01

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

  13. Compact torsatron reactors

    International Nuclear Information System (INIS)

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  14. Multi-purpose reactor

    International Nuclear Information System (INIS)

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MWth, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co60) production capacity is 50000 Ci/yr, 200Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  15. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author)

  16. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  17. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  18. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  19. Oxide-metal cores - stage of conversion to the metal fuel core for the fast reactors of the BN-type

    International Nuclear Information System (INIS)

    Radiation-thermal properties of uranium-containing metal fuel were studied, that influence reliability parameters of blanket and fuel elements for BN reactors. Specifically, the following was investigated: macroeffects of radiation growth of metal fuel column in steel claddings; macroeffects of radiation swelling of uranium and U alloys in free states and steel claddings; and also macroeffects of physical chemical interaction between uranium/uranium-plutonium alloys and steel claddings. Dependences of these macroeffects were established on types and mass content of doped additives (0,003...40 wt.% Me, MeO...), on irradiation temperature (TT=100..900 deg. C), burnup factor (B ≤ 10,4% h.a.), burnup rate (1· 1013fissions./cm3·sec ≤ ω ≤ 12·1013fissions./cm3·sec), design strength of steel claddings (δ/d = 0,03...0,11), cladding temperature under irradiation (TcladMax ≤ 750 deg. C) and postirradiation emergency overheatings TTMax = TcladMax ≤ 900 deg. C), smear fuel density in fuel mockups and full-sized elements (γeff=12...18 gh.a./cm3), on types and thickness of antidiffusion protective layers on the border between metal fuel column and steel cladding (nonmetal and metal layers of 5...40μm in thickness). The established dependences were used for development and fabrication of experimental metal uranium fuel pins that were intended for operational conditions and parameters in heterogeneous oxide-metal cores of various types: BFAH (by FA heterogenization), IFAH (intra FA heterogenization), IFPH (intra fuel pin heterogenization). Main features of the fuel pins: - claddings- made of austenitic steel EI-847 of standard type-dimensions: dxδ= 6,0x0,3mm; 6,9x0,4 (0,5)mm; 14,5x0,45mm; - filling of the fuel pins - He; - fuel columns - alloy-free uranium (depleted, natural, enriched, with additives of up to 8%wt plutonium); smear density of fuel - 12...18 gh.a./cm3, design similarity to the standard oxide fuel pins applied in the BOR-60, BN-350, BN-600 reactors

  20. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  1. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

  2. Safety of research reactors

    International Nuclear Information System (INIS)

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

  3. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  4. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  5. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  6. FBR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an FBR type reactor in which the combustion of reactor core fuels is controlled by reflectors, and the position of a reflector driving device can be controlled even during shut down of the reactor. Namely, the reflector driving device is attracted to the outer wall surface of a reactor core barrel by electromagnetic attraction force. An inertia body is disposed vertically movably to the upper portion of the reflector driving device. Magnetic repulsive coils generate instantaneous magnetic repulsive force between the inertia body and the reflector driving device. With such a constitution, the reflector driving device can be driven by using magnetic repulsion of the electromagnetic repulsive coils and inertia of the inertia body. As a result, not only the reflectors can be elevated at an ultraslow speed during normal reactor operation, but also fine position adjustment for the reflector driving device, as well as fine position adjustment of the reflectors required upon restart of the reactor can be conducted by lowering the reflector driving device during shut down of the reactor. (I.S.)

  7. Reactor water sampling device

    International Nuclear Information System (INIS)

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  8. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  9. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

  10. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

    The device of the present invention monitors phenomena occurred in a nuclear reactor more accurately than usual case. that is, the device monitors a reactor power by signals sent from a great number of neutron monitors disposed in the reactor. The device has a means for estimating a phenomenon occurred in the reactor based on the relationship of a difference of signals between each of the great number of neutron monitors to the positions of the neutron monitors disposed in the reactor. The estimation of the phenomena is conducted by, for example, conversion of signals sent from the neutron monitors to a code train. Then, a phenomenon is estimated rapidly by matching the code train described above with a code train contained in a data base. Further. signals sent from the neutron monitors are processed statistically to estimate long term and periodical phenomena. As a result, phenomena occurred in the reactor are monitored more accurately than usual case, thereby enabling to improve reactor safety and operationability. (I.S.)

  11. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  12. Determination of research reactor safety parameters by reactor calculations

    International Nuclear Information System (INIS)

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

  13. Reactor de plasma

    OpenAIRE

    Erra Serrabasa, Pilar; Molina Mansilla, Ricardo; Beltrán Serra, Eric

    2008-01-01

    Reactor de plasma. Se trata de un reactor de plasma que puede trabajar en un amplio rango de presión, desde el vacío y presiones reducidas hasta la presión atmosférica y presiones superiores. Adicionalmente el reactor de plasma tiene la capacidad de regular otros parámetros importantes y permite su uso para el tratamiento de muestras de tipología muy diversa, como por ejemplo las de tamaño relativamente grande o de superficie rugosa.

  14. Integral nuclear reactor

    International Nuclear Information System (INIS)

    The invention deals with an inprovement of the design of an integral pressurized water nuclear reactor. A typical embodyment of the invention includes a generally cylindrical pressure vessel that is assembled from three segments which are bolted together at transverse joints to form a pressure tight unit that encloses the steam generator and the reactor. The new construction permits primary to secondary coolant heat exchange and improved control rod drive mecanisms which can be exposed for full service access during reactor core refueling, maintenance and inspection

  15. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  16. Licensed operating reactors

    International Nuclear Information System (INIS)

    The Operating Units Status Report --- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff on NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- power reactors in the US

  17. First Algerian research reactor

    International Nuclear Information System (INIS)

    In 1985, both the Algerian Commissariat of New Energies and the Argentine National Atomic Energy Commission plus the firm INVAP S.E., started a series of mutual visits aimed at defining the mechanisms for cooperation in the nuclear field. Within this framework, a commercial contract was undersigned covering the supply of a low-power reactor (RUN), designed for basic and applied research in the fields of reactor physics and nuclear engineering. The reactor may also be used for performing experiences with neutron beams, for the irradiation of several materials and for the training of technicians, scientists and operators

  18. Course on reactor physics

    International Nuclear Information System (INIS)

    In Germany only few students graduate in nuclear technology, therefore the NPP operating companies are forced to develop their own education and training concepts. AREVA NP has started together with the Technical University of Dresden a one-week course ''reactor physics'' that includes the know-how of the nuclear power plant construction company. The Technical University of Dresden has the training reactor AKR-2 that is retrofitted by modern digital instrumentation and control technology that allows the practical training of reactor control.

  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. Nuclear reactor theory

    International Nuclear Information System (INIS)

    This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

  1. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  2. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

  3. International tokamak reactor

    International Nuclear Information System (INIS)

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  4. 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)

  5. 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.

  6. nuclear reactor design calculations

    International Nuclear Information System (INIS)

    In this work , the sensitivity of different reactor calculation methods, and the effect of different assumptions and/or approximation are evaluated . A new concept named error map is developed to determine the relative importance of different factors affecting the accuracy of calculations. To achieve this goal a generalized, multigroup, multi dimension code UAR-DEPLETION is developed to calculate the spatial distribution of neutron flux, effective multiplication factor and the spatial composition of a reactor core for a period of time and for specified reactor operating conditions. The code also investigates the fuel management strategies and policies for the entire fuel cycle to meet the constraints of material and operating limitations

  7. Nuclear reactor internal structures

    International Nuclear Information System (INIS)

    The upper internal structures of the reactor are connected to the closing head so as to be readily removed with the latter and a skirt connected to the lower portion of said upper structures so as to surround the latter, extends under the control rods when they are removed from the reactor core. Through such an arrangement the skirt protects the control rods and supports the vessel closing-head and the core upper structures, whenever the head is severed from the vessel and put beside the latter in order to discharge the reactor

  8. Reactor monitoring system

    International Nuclear Information System (INIS)

    The present invention concerns a device for monitoring the inside of an FBR type reactor which can not be monitored by a usual optical camera. An ultrasonic camera having an excellent propagating property in a liquid metal sodium is scanned, and reflected waves of the ultrasonic waves are received as signals. The signals are processed by using a virtual realistic feeling (VR) technique such as a head mounting type image display (HMD) and a three dimensional pointing device. With such procedures, the inside of the FBR type reactor can be observed with such a realistic feeling that the inside of the FBR type reactor were seen directly. (I.S.)

  9. Research reactor support

    International Nuclear Information System (INIS)

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  10. Study of power reactor dynamics by stochastic reactor oscillator method

    International Nuclear Information System (INIS)

    Stochastic reactor oscillator and cross correlation method were used for determining reactor dynamics characteristics. Experimental equipment, fast reactor oscillator (BOR-1) was activated by random pulses from the GBS-16 generator. Tape recorder AMPEX-SF-300 and data acquisition tool registered reactor response to perturbations having different frequencies. Reactor response and activation signals were cross correlated by digital computer for different positions of stochastic oscillator and ionization chamber

  11. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  12. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention concerns a method of controlling reactor power to shift it into a partial power operation upon occurrence of recycling pump tripping or loss of generator load. Operation state of a reactor is classified into a plurality of operation states based on values of the reactor core flow rate and the reactor power. Different insertion patterns for selected control rods are determined on every classified operation states. Then, an insertion pattern corresponding to the operation state upon occurrence of recycling pump tripping or loss of power generator load is carried out to shift into partial power operation. The operation is shifted to a load operation solely in the station while avoiding risks such as TPM scram. Then neutron fluxes are suppressed upon transient to increase margin of fuel integrity. Selected control rod pattern of the optimum reactivity is set to each of operation regions, thereby enabling to conduct flexible countermeasure so as to attain optimum operationability. (N.H.)

  13. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  14. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  15. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  16. Reactor parameter simulation system

    International Nuclear Information System (INIS)

    A reactor parameter simulation system (RPSS) has been built with the capability of analyzing any reactor signals, decomposing those signals into their deterministic and stochastic components, then reconstructing new, simulated signals that possess the same statistical and correlation structure as the original plant variables. Important uses of the RPSS are for integration with reactor simulation software to provide tools for plant control strategy development, and for safety-study investigations of scenarios that can arise involving signal faults generated from degraded sensors. A third use of the RPSS is for frequency-domain filtering of reactor process variables contaminated with serially correlated noise, which is important for our ongoing development of expert systems for sensor-operability surveillance. 5 refs., 4 figs., 3 tabs

  17. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  18. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  19. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  20. Ageing of research reactors

    International Nuclear Information System (INIS)

    Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

  1. Experience with Kamini reactor

    International Nuclear Information System (INIS)

    Kamini is a 233U fuelled, 30 kW(th) research reactor. It is one of the best neutron source facility with a core average flux of 1012 n/cm2/s in IGCAR used for neutron radiography of active and nonradioactive objects, activation analysis and radiation physics research. The core consists of nine plate type fuel elements with a total fuel inventory of 590 g of 233U. Two safety control plates made of cadmium are used for start up and shutdown of the reactor. Three beam tubes, two-thimble irradiation site outside reflector and one irradiation site nearer to the core constitute the testing facilities of Kamini. Kamini attained first criticality on 29th October 96 and nominal power of 30 kW in September 1997. This paper covers the design features of the reactor, irradiation facilities and their utilities and operating experience of the reactor. (author)

  2. Dossier: research reactors

    International Nuclear Information System (INIS)

    Research reactors are used at the CEA (the French atomic energy commission) since many years. Their number has been reduced but they remain unique tools that CEA valorize continuously. The results of the programs involving such reactors are of prime importance for the operation of Electricite de France (EdF) park of existing power plants but also for the design of future nuclear power plants and future research reactors. This dossier presents three examples of research reactors in use at the CEA: Osiris and Orphee (CEA-Saclay), devoted to nuclear energy and fundamental research, respectively, and the critical mockups Eole, Minerve and Masurca (CEA-Cadarache) devoted to nuclear data libraries and neutronic calculation. (J.S.)

  3. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  4. Future Reactor Experiments

    CERN Document Server

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measurement techniques have been explored. A proposed experiment JUNO, with a 20 kton liquid scintillator detector of $3%/$$\\sqrt{E(MeV)}$ energy resolution, $\\sim$ 53 km far from reactors of $\\sim$ 36 GW total thermal power, can reach to a sensitivity of $\\Delta\\chi^{2}>16$ considering the spread of reactor cores and uncertainties of the detector response. Three of mixing parameters are expected to be measured to better than 1% precision. There are multiple detector options for JUNO under investigation. The technical challenges...

  5. Reactor hot spot analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R.B.

    1985-08-01

    The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

  6. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  7. Research Reactor Benchmarks

    International Nuclear Information System (INIS)

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given

  8. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author)

  9. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  10. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

  11. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  12. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

  13. Special lecture on nuclear reactor

    International Nuclear Information System (INIS)

    This book gives a special lecture on nuclear reactor, which is divided into two parts. The first part has explanation on nuclear design of nuclear reactor and analysis of core with theories of integral transports, diffusion Nodal, transports Nodal and Monte Carlo skill parallel computer and nuclear calculation and speciality of transmutation reactor. The second part deals with speciality of nuclear reactor and control with nonlinear stabilization of nuclear reactor, nonlinear control of nuclear reactor, neural network and control of nuclear reactor, control theory of observer and analysis method of Adomian.

  14. The replacement research reactor

    International Nuclear Information System (INIS)

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  15. OECD Halden reactor project

    International Nuclear Information System (INIS)

    This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

  16. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  17. Small reactor return

    International Nuclear Information System (INIS)

    Current state of the development of present-day small reactors in different countries is performed. Various designs of low and middle power reactors, among which are CAREM (25 MW, PWR), KLT-40 (40 MW, PWR), MRX (30 MW, PWR), IRIS (50 MW, PWR), SMART (1000 MW, PWR), Modular SBWR (50 MW, BWR), PBMR (120 MW, HTGR), GT-HMR (285 MW, HTGR), are discussed

  18. Reactor lattice transport calculations

    International Nuclear Information System (INIS)

    The present lecture is a continuation of the lecture on Introduction to the Neutron Transport Phenomena. It comprises three aspects of lattice calculations. First the idea of a reactor lattice is introduced. Then the main definitions used in reactor lattice analysis are given, and finally two basic methods applied for solution of the transport equations are defined. Several remarks on secondary results from lattice transport calculations are added. (author)

  19. Thermal or epithermal reactor

    International Nuclear Information System (INIS)

    In a thermal or epithermal heavy-water reactor of the pressure tube design the reactivity is to be increased by different means: replacement of the moderator by additional rods with heavy metal in the core or in the reflector; separation of the moderator (heavy water) from the coolant (light water) by means of shroud tubes. In light-water reactor types neutron losses are to be influenced by using the heavy elements in different configurations. (orig./PW)

  20. Future reactor experiments

    International Nuclear Information System (INIS)

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper

  1. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  2. Jet-Stirred Reactors

    OpenAIRE

    Herbinet, Olivier; Guillaume, Dayma

    2013-01-01

    The jet-stirred reactor is a type of ideal continuously stirred-tank reactor which is well suited for gas phase kinetic studies. It is mainly used to study the oxidation and the pyrolysis of hydrocarbon and oxygenated fuels. These studies consist in recording the evolution of the conversion of the reactants and of the mole fractions of reaction products as a function of different parameters such as reaction temperature, residence time, pressure and composition of the inlet gas. Gas chromatogr...

  3. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

  4. Future reactor experiments

    Science.gov (United States)

    Wen, Liangjian

    2015-07-01

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  5. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  6. AVR reactor physics

    International Nuclear Information System (INIS)

    A process for reactivity control was developed and used for fuelling the AVR reactor core, which is largely based on experimentally determined values. By adding fuel elements with different quantities of heavy metals paired with various experimental requirements, great demands were made of reactivity control. Although only a small range of control was available, this was sufficient to operate the reactor and to shut it down safely in the required power and temperature range. (orig.)

  7. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  8. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  9. Emergency reactor scram system

    International Nuclear Information System (INIS)

    The present invention provides an emergency reactor scram system capable of shut down a reactor safely upon occurrence of pump trip by improving a passive scram performance for an FBR-type reactor. Namely, a driving motor and an electric generator are connected to a main pump of a primary system. An AC/DC convertor is connected to the electric generator. A shielding plug is disposed to the upper end opening of a reactor container, a control rod drive mechanism is erected on the shielding plug, and an extension pipe is attached to scram magnets of the control rod drive mechanism. The extension pipe is connected to a control rod. The rotation of the shaft of the pump is used as a direct rotator to provide an integrated-type electric generator. The electric generator is electrically connected with the power source of a scram magnet of the emergency scram system. Accordingly, the control rod of the emergency scram system is automatically and rapidly inserted to the reactor core using the power source of the electric generator upon trip of the main pump thereby enabling to scram the reactor safely. (I.S.)

  10. A modular reactor plant

    International Nuclear Information System (INIS)

    This paper describes a new concept in liquid metal reactors that is being developed by General Electric under contract to the Department of Energy. This concept is called the Modular Reactor Plant. While this effort is not expected to have a near-term impact, it is directed toward three principal issues currently affecting nuclear power in the United States. First, plant costs have escalated to the point where the startup of new plants require large electric rate increases. Second, the cost of new plants coming on-line today vary by as much as a factor of three. And, third, nuclear construction times often exceed the utilities prudent planning cycle. This paper describes how General Electric's Modular Reactor Plant addreses these issues through shop fabrication and assembly, rail shipment to the site for rapid installation of nuclear components and inherent reactor protection. In addition, it is expected the modular reactor plant will reduce the current cost of development and demonstration of liquid metal reactors to an affordable level

  11. New fission reactor designs

    International Nuclear Information System (INIS)

    A number of critical challenges to the expanded or continued use of nuclear power have developed. These can be categorized as: regulatory restrictions and complications; negative public attitudes; plant complexity; plant life, operations, and maintenance; uncertain load growth, financing; waste management. Solutions to these challenges through advanced reactor design centre around four key technical responses. Passive safety systems are being introduced which use the laws of physics to provide emergency reactor coding, control and shutdown thus eliminating the possibility of human error. Modular construction promises cuts in costs and construction time by shifting the major part of component manufacture from the site to the factory. Standardization also cuts capital costs and in addition operations and repair costs and expedites reactor licensing. Improvements to the fuel cycle include improved fuel types, designs and fabrication, and the reprocessing of and recycling spent fuel back into energy production, thus extending uranium resources and offering a partial solution to the problem of waste disposal. Examples of evolutionary and advanced water-cooled reactors, modular high temperature gas-cooled reactors, and advanced liquid metal cooled fast breeder reactors which are being developed round the world are presented. (author)

  12. OECD Halden reactor project

    International Nuclear Information System (INIS)

    This is the nineteenth annual Report on the OECD Halden Reactor Project, describing activities at the Project during 1978, the last year of the 1976-1978 Halden Agreement. Work continued in two main fields: test fuel irradiation and fuel research, and computer-based process supervision and control. Project research on water reactor fuel focusses on various aspects of fuel behavior under normal, and off-normal transient conditions. In 1978, participating organisations continued to submit test fuel for irradiation in the Halden boiling heavy-water reactor, in instrumented test assemblies designed and manufactured by the Project. Work included analysis of the impact of fuel design and reactor operating conditions on fuel cladding behavior. Fuel performance modelling included characterization of thermal and mechanical behavior at high burn-up, of fuel failure modes, and improvement of data qualification procedures to reduce and quantify error bands on in-reactor measurements. Instrument development yielded new or improved designs for measuring rod temperature, internal pressure, axial neutron flux shape determination, and for detecting cladding defects. Work on computer-based methods of reactor supervision and control included continued development of a system for predictive core surveillance, and of special mathematical methods for core power distribution control

  13. Reactor power measuring device

    International Nuclear Information System (INIS)

    The device of the present invention efficiently calibrates a fixed type gamma ray thermometer of a reactor power measuring device of a BWR type reactor. Namely, the device of the present invention calculates peripheral fuel rod power distribution by calibrating the reactor power distribution by heat generation amount, the reactor power distribution being obtained by a calculation based on a reactor model for converting the signals of a plurality of the gamma ray thermometers in the reactor core based on a conversion formula. In this case, the conversion formula is a relational formula between the power of a thermocouple of the gamma ray thermometer, gamma ray heat generation amount, thermocouple zero power sensitivity relative to a temperature coefficient. A conversion efficient calculation means makes a calibration heater to generate heat at a predetermined power, and the thermocouple zero power sensitivity and the temperature coefficient are obtained based on the output of the gamma ray thermometer in this case. The calibration means updates to conversion type thermocouple zero power sensitivity and temperature coefficient. A calibration execution means executes the operations described above successively, and when the thermocouple zero power sensitivity and the temperature coefficient are out of an allowable range, the means informs it and eliminates the corresponding gamma ray thermometer from the measuring meters. (I.S.)

  14. Reactor safety engineering

    International Nuclear Information System (INIS)

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

  15. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions

  16. Experimental studies on acoustic detection of sodium-water steam generator leaks in the USSR

    International Nuclear Information System (INIS)

    The paper reports that the acoustic leak indicators have been developed in two versions. The first one is based upon using the immersible acoustic hydrophones and the parallel frequency analysis of their signals. The second one uses the waveguide sensors with microprocessor system of noise signals processing. Brief description of both versions is given. The result of these systems tests at the experimental facilities, BN-600 and BOR-60 reactors are also provided. 4 refs, 15 figs

  17. The reactor Cabri

    International Nuclear Information System (INIS)

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m3/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

  18. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  19. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  20. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

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

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

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

  2. 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

  3. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3He, 6Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  4. Reactor coolant cleanup facility

    International Nuclear Information System (INIS)

    A depressurization device is disposed in pipelines upstream of recycling pumps of a reactor coolant cleanup facility to reduce a pressure between the pressurization device and the recycling pump at the downstream, thereby enabling high pressure coolant injection from other systems by way of the recycling pumps. Upon emergency, the recycling pumps of the coolant cleanup facility can be used in common to an emergency reactor core cooling facility and a reactor shutdown facility. Since existent pumps of the emergency reactor core cooling facility and the reactor shutdown facility which are usually in a stand-by state can be removed, operation confirmation test and maintenance for equipments in both of facilities can be saved, so that maintenance and reliability of the plant are improved and burdens on operators can also be mitigated. Moreover, low pressure design can be adopted for a non-regenerative heat exchanger and recycling coolant pumps, which enables to improve the reliability and economical property due to reduction of possibility of leakage. (N.H.)

  5. HTGR type reactor

    International Nuclear Information System (INIS)

    A reactor core is disposed at the center of a reactor container, a reflector is disposed on the outer side thereof, a steam generator is disposed further outer side thereof coaxially, and they are constituted as an integrated one container. A gas circulator and control rod drives are protruded at the outer side of the lower portion of the integrated container. Heat insulators are disposed on the inner side of the container wall in the upper portion of the reactor container. Helium gas risen in the reactor core and heated to a high temperature descends in a circular steam generator and undergoes heat exchange with water, and is then pressurized in the gas circulator after the lowering of the temperature, and returned to the inlet of the reactor core from the lower central portion of the container. With such procedures, the helium gas as primary coolants circulates only in the container to improve confinement. The device can be reduced in the size and the cost. (I.N.)

  6. Reactor container spray device

    International Nuclear Information System (INIS)

    Purpose: To enable decrease in the heat and the concentration of radioactive iodine released from the reactor vessel into the reactor container in the spray device of BWR type reactors. Constitution: A plurality of water receiving trays are disposed below the spray nozzle in the dry well and communicated to a pressure suppression chamber by way of drain pipeways passing through a diaphragm floor. When the recycling system is ruptured and coolants in the reactor vessel and radioactive iodine in the reactor core are released into the dry well, spray water is discharged from the spray nozzle to eliminate the heat and the radioactive iodine in the dry well. In this case, the receiving trays collect the portions of spray water whose absorption power for the heat and radioactive iodine is nearly saturated and falls them into the pool water of the pressure suppression chamber. Consequently, other portions of the spray water that still possess absorption power can be jetted with no hindrance, to increase the efficiency for the removal of the heat and iodine of the spray droplets. (Horiuchi, T.)

  7. PROTEUS research reactor

    International Nuclear Information System (INIS)

    The PROTEUS zero power reactor at the Paul Scherrer Institute (PSI) in Switzerland achieved first criticality in 1968 and since then has been operated as an experimental tool for reactor physics research on test lattices representative of a wide range of reactor concepts. Reactor design codes and their associated data libraries are validated on the basis of the experimental results obtained. PROTEUS is normally configured as a driven system, in which a subcritical test zone is made critical by the surrounding driver zones. The advantages of driven systems can be summarized as follows: - Smaller amount of test fuel is required; - Large range of test zone conditions (including k∞ < 1 states) can be investigated by changes in the driver loading alone, thus avoiding undesirable perturbations to the test zone which would influence the measurement conditions and thus affect the interpretability of the results; - Necessary reactor control and instrumentation equipment (usually perturbing from the experimental viewpoint) can be located in the outer driver regions, thereby avoiding disturbance of the test lattice

  8. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  9. 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.)

  10. EBT reactor analysis

    International Nuclear Information System (INIS)

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this operating window, physics and engineering systems analysis and cost sensitivity studies indicate that reactors with approx. 6 to 10%, P approx. 1200 to 1700 MW(e), wall loading approx. 1.0 to 2.5 MW/m2, and recirculating power fraction (including ring-sustaining power and all other reactors auxiliaries) approx. 10 to 15% are possible. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters. These include, but are not limited to, the use of: (1) supplementary coils or noncircular mirror coils to improve magnetic geometry and reduce size, (2) energetic ion rings to improve ring power requirements, (3) positive potential to enhance confinement and reduce size, and (4) profile control to improve stability and overall fusion power density

  11. Modern research reactors in the world and RA research reactor

    International Nuclear Information System (INIS)

    This paper covers the following topics: fundamentals of research reactors, thermal neutron flux density, classification of research reactors in the world, properties of research reactors of higher power in the world according to IAEA data for 1995, their application, and trend of development, experimental feasibility and status of RA reactor. Trend of research reactors development in the world (after 1980) is directed towards increasing the neutron production quality factor, i.e. ratio between thermal neutron flux density and reactor power, which is achieved by designing compact reactor cores. With the aim of renewal of RA reactor (without analysis of reactor components and staff aging, possibility of restart and commercialization), according to the analysis in this paper, it can be concluded: there is very few reactors under construction in the world, all the important countries in Europe have research reactors; RA reactor is not very interesting for development of reactor physics; nowadays RA reactor is in the group of reactors which are 30-40 years old; its inventories of fuel and heavy water are enough for about 20 years of operation; it has achieved high quality factor of neutron production with low and highly enriched fuel; core transfer from low highly enriched to low enriched fuel should be carefully studies from operation, experimental and economical point of view; it is necessary to use the advantages of RA reactor (minimum investment): volume of the core and reflector which enables availability of neutron flux for the users (numerous experimental loops), fuel in shape of slugs enabling efficient fuel management and flexible neutron flux distribution in the core in the reflector, reactor operation should be directed towards commercial applications. Bibliography of more than 140 relevant papers used is included in this paper

  12. Sodium-cooled nuclear reactors

    International Nuclear Information System (INIS)

    This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index

  13. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  14. Methanation assembly using multiple reactors

    Science.gov (United States)

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  15. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  16. MINT research reactor safety program

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

  17. RB research reactor Safety Report

    International Nuclear Information System (INIS)

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

  18. Fusion reactor materials

    International Nuclear Information System (INIS)

    This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

  19. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

  20. 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

  1. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  2. Licensed operating reactors

    International Nuclear Information System (INIS)

    THE OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

  3. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  4. Colliding Beam Fusion Reactors

    Science.gov (United States)

    Rostoker, Norman; Qerushi, Artan; Binderbauer, Michl

    2003-06-01

    The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker-Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are generic in that they do not relate to specific confinement devices. In all cases except for a Tokamak with D-T fuel the recirculating power was found to exceed the fusion power by a large factor. In this paper we criticize the generality claimed for this calculation. The ratio of circulating power to fusion power is calculated for the Colliding Beam Reactor with fuels D-T, D-He3 and p-B11. The results are respectively, 0.070, 0.141 and 0.493.

  5. The MNSR reactor

    International Nuclear Information System (INIS)

    This tank-in-pool reactor is based on the same design concept as the Canadian Slowpoke. The core is a right circular cylinder, 24 cm diameter by 25 cm long, containing 411 fuel pin positions. The pins are HEU-Aluminium alloy, 0.5 cm in diameter. Critical mass is about 900 g. The reactor has a single cadmium control rod. The back-up shutdown system is the insertion of a cadmium capsule in a core position. Excess reactivity is limited to 3.5mk. In both the MNSR and Slowpoke, the insertion of the maximum excess reactivity results in a power transient limited by the coolant/moderator temperature to safe values, independent of any operator action. This reactor is used primarily in training and neutron activation analysis. Up to 64 elements have been analyzed in a great variety of different disciplines. (author)

  6. Welding and reactor safety

    International Nuclear Information System (INIS)

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

  7. Reactor operation experience

    International Nuclear Information System (INIS)

    Since the TRIGA Users Conference in Helsinki 1970 the TRIGA reactor Vienna was in operation without any larger undesired shutdown. The integrated thermal power production by August 15 1972 accumulated to 110 MWd. The TRIGA reactor is manly used for training of students, for scientific courses and research work. Cooperation with industry increased in the last two years either in form of research or in performing training courses. Close cooperation is also maintained with the IAEA, samples are irradiated and courses on various fields are arranged. Maintenance work was performed on the heat exchanger and to replace the shim rod magnet. With the view on the future power upgrading nine fuel elements type 110 have been ordered recently. Experiments, performed currently on the reactor are presented in details

  8. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  9. Reactor accidents in perspective

    International Nuclear Information System (INIS)

    In each of the three major reactor accidents which have led to significant releases to the environment, and discussed in outline in this note, the reactor has been essentially destroyed - certainly Windscale and Chernobyl reactors will never operate and the cleanup operation for Three Mile Island is currently estimated to have cost in excess of US Pound 500 000 000. In each of the accidents there has not been any fatality off site in the short term and any long-term health detriment is unlikely to be seen in comparison with the natural cancer incidence rate. At Chernobyl, early fatalities did occur amongst those concerned with fighting the incident on site and late effects are to be expected. The assumption of a linear non-threshold risk, and hence no level of zero risk is the main problem in communication with the public, and the author calls for simplification of the presentation of the concepts of radiological protection. (U.K.)

  10. Reactor safety equipments

    International Nuclear Information System (INIS)

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

  11. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  12. Reactor protection system

    International Nuclear Information System (INIS)

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  13. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  14. MERCHANT MARINE SHIP REACTOR

    Science.gov (United States)

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

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

    International Nuclear Information System (INIS)

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

  16. Study of future reactors

    International Nuclear Information System (INIS)

    Today, more than 420 large reactors with a gross output of close to 350 GWe supply 20 percent of world electricity needs, accounting for less than 5 percent of primary energy consumption. These figures are not expected to change in the near future, due to suspended reactor construction in many countries. Nevertheless, world energy needs continue to grow: the planet's population already exceeds five billion and is forecast to reach ten billion by the middle of the next century. Most less developed countries have a very low rate of energy consumption and, even though some savings can be made in industrialized countries, it will become increasingly difficult to satisfy needs using fossil fuels only. Furthermore, there has been no recent breakthrough in the energy landscape. The physical feasibility of the other great hope of nuclear energy, fusion, has yet to be proved; once this has been done, it will be necessary to solve technological problems and to assess economic viability. Although it is more ever necessary to pursue fusion programs, there is little likelihood of industrial applications being achieved in the coming decades. Coal and fission are the only ways to produce massive amounts of energy for the next century. Coal must overcome the pollution problems inherent in its use; fission nuclear power has to gain better public acceptance, which is obviously colored by safety and waste concerns. Most existing reactors were commissioned in the 1970s; reactor lifetime is a parameter that has not been clearly established. It will certainly be possible to refurbish some to extend their operation beyond the initial target of 30 or 40 years. But normal advances in technology and safety requirements will make the operation of the oldest reactors increasingly difficult. It becomes necessary to develop new generations of nuclear reactors, both to replace older ones and to revive plant construction in their countries that are not yet equipped or that have halted their

  17. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  18. Operating US power reactors

    International Nuclear Information System (INIS)

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of Dec. 31, 1986, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (October, November, and December 1986) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. They are defined as follows: In addition to the tabular data, this article discusses significant occurrences and developments that affected licensed US power reactors during this reporting period. It includes, but is not limited to, changes in operating status, regulatory actions and decisions, and legal actions involving the status of power reactors. We do not have space here for routine problems of operation and maintenance, but such information is available at the Nuclear Regulatory Commission (NRC) Public Document Room, 1717 H Street, NW, Washington, DC 20555. Some significant operating events are summarized elsewhere in this section in the article ''Selected Safety-Related Events,'' and a report on activities relating to facilities still in the construction process is given in the article ''Status of Power-Reactor Projects Undergoing Licensing Review'' in the last section of each issue of this journal. The reader's attention is also called to the regular feature ''General Administrative Activities,'' which deals with more general aspects of regulatory and legal matters that are not covered elsewhere in the journal

  19. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  20. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  1. 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

  2. Diagnostics for hybrid reactors

    International Nuclear Information System (INIS)

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  3. Perspectives on reactor safety

    International Nuclear Information System (INIS)

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

  4. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

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

  5. Netherlands Interuniversity Reactor Institut

    International Nuclear Information System (INIS)

    This is the annual report of the Interuniversity Reactor Institute in the Netherlands for the Academic Year 1977-78. Activities of the general committee, the daily committee and the scientific advice board are presented. Detailed reports of the scientific studies performed are given under five subjects - radiation physics, reactor physics, radiation chemistry, radiochemistry and radiation hygiene and dosimetry. Summarised reports of the various industrial groups are also presented. Training and education, publications and reports, courses, visits and cooperation with other institutes in the area of scientific research are mentioned. (C.F.)

  6. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  7. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

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

  8. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  9. Decay of reactor neutrinos

    OpenAIRE

    Vogel, P.

    1984-01-01

    We consider the decay of massive neutrinos which couple to electrons and are, therefore, produced in nuclear reactors. Lifetime limits for the γ and electron-positron decay modes of these neutrinos are deduced from the experimental limit on the singles count rate in the detector used to study neutrino oscillations at the Gösgen reactor. The dominantly coupled neutrinos are light, and their invariant-lifetime limit tc.m. / mν is 1-3 sec/eV. The subdominantly coupled heavy neutrinos with mass 1...

  10. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

  11. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  12. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  13. Reactor PIK construction

    International Nuclear Information System (INIS)

    The construction work at the 100 MW researches reactor PIK in year 2002 was in progress. The main activity was concentrated on mechanical, ventilation and electrical equipment. Some systems and subsystems are under adjustment. Hydraulic driving gear for beam shutters are finished in installation, rinsing, and adjusting. Regulating rods test assembling was done. On the critical assembly the first reactor fueling was tested to evaluate the starting neutron source intensity and a sufficiency of existing control and instrument board. Mainline of the PIK facility design and neutron parameters are presented. (author)

  14. Reactor pressure vessel materials

    International Nuclear Information System (INIS)

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 3 offers a detailed treatment of the selection criteria and properties of reactor pressure vessel materials. The main attention is directed towards steel and ingot making and the subsequent material processing

  15. Reactor gamma spectrometry: status

    International Nuclear Information System (INIS)

    Current work is described for Compton Recoil Gamma-Ray Spectrometry including developments in experimental technique as well as recent reactor spectrometry measurements. The current status of the method is described concerning gamma spectromoetry probe design and response characteristics. Emphasis is given to gamma spectrometry work in US LWR and BR programs. Gamma spectrometry in BR environments are outlined by focussing on start-up plans for the Fast Test Reactor (FTR). Gamma spectrometry results are presented for a LWR pressure vessel mockup in the Poolside Critical Assembly (PCA) at Oak Ridge National Laboratory

  16. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

  17. Space-time reactor kinetics for heterogeneous reactor structure

    International Nuclear Information System (INIS)

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods

  18. Risk prevention during reactor shutdown

    International Nuclear Information System (INIS)

    During reactor shutdown potential risks are issued of a number of maintenance operations. In this text we analyse these operations and give the modifications of technical specifications to ameliorate the reactor safety. 4 figs

  19. 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

  20. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  1. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  2. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  3. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  4. The IR-8 reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ryazantsev, E.P.; Egorenkov, P.M.; Yashin, A.F. [Reactor Technology and Materials Research Inst. of RRC ' KI' , Moscow (Russian Federation)

    1997-07-01

    At the Russian Research Center 'Kurchatov Institute' (RRC 'KI') the IR-8 reactor commissioning was carried out in 1981. The reactor was developed in return for earlier existing at RRC 'KI' of the IRT-M reactor (modernized IRT reactor, constructed in 1957). The IRT-M reactor was used for investigations in nuclear physics, solid state physics, radiation chemistry, biology as well as to produce isotopes. Under developing the IR-8 reactor the IRT biological shielding with beam tubes and its process systems were used. The IR-8 reactor creation was founded on application developed by then new fuel assemblies (FA) of IRT-3M type, having two times as great surface of heat transfer and 1.75 times higher U-235 load than the FA of the IRT-2M type, which were used in IRT-M reactor. (author)

  5. Power calibrations for TRIGA reactors

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than ± 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to ± 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  6. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

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

  7. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

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

  8. VVER and RBMK reactors

    International Nuclear Information System (INIS)

    The safety of VVER and RBMK reactors has been discussed a lot after Chernobyl accident. Some improvements have been performed since that especially in RBMK-reactors and extensive programmes for backfitting have been planned and are partly underway. There are two different sizes of VVER reactors, 440 MW and 1000 MW. The design bases and designs itself vary inside the family of two size classes depending on the age of the plant. The oldest VVER-440 is called model 230 and the newest model 213. The oldest VVER-1000 units (two units) are prototypes that have some unique, nonfavorable features. The next stage of VVER-1000 developement (three units) is model V-302 and the remaining 15 plants in operation are model V-320, but even within this latest model there are some differences. The design bases and designs vary also inside the family of the RBMK reactors exactly the same way as in VVERs. The most important design bases of nuclear power plants designed in the former Soviet Union is presented in this paper. Also some safety advantages and disadvantages of these NPPs are discussed. (au). (5 figs.)

  9. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

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

  10. Studies on reactor physics

    International Nuclear Information System (INIS)

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  11. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  12. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  13. Thermal Reactor Safety

    International Nuclear Information System (INIS)

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

  14. Nuclear reactor container

    International Nuclear Information System (INIS)

    In a container of a BWR type reactor, spray water is stored in a pedestal cavity. A perforated hole is formed on the side wall of the pedestal, and a stirrer is disposed in the pedestal cavity to stir the stored spray water. During reactor operation, the door on the side wall of the pedestal is closed to prevent discharge of fission products to the dry well when a severe accident should occur. During periodical inspection for the plant, the door is opened to enable an operator to access to the inside of the pedestal. When a molten reactor core should drop to the pedestal cavity, fission products generated from the failed reactor core left in a pressure vessel pass through the spray water in the pedestal cavity. Then, most of the fission products are held in the spray water by a scrubbing effect when they pass through the spray water. In addition, the stored spray water is stirred by the stirrer to enhance the scrubbing effect thereby enabling to further decrease the amount of the fission products discharged to the dry well. (N.H.)

  15. ICF tritium production reactor

    International Nuclear Information System (INIS)

    The conceptual design of an ICF tritium production reactor is described. The chamber design uses a beryllium multiplier and a liquid lithium breeder to achieve a tritium breeding ratio of 2.08. The annual net tritium production of this 532 MW/sub t/ plant is 16.9 kg, and the estimated cost of tritium is $8100/g

  16. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  17. Nuclear reactor building

    Science.gov (United States)

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  18. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A method of constructing a radiation shielding plug for use in the roof of the coolant containment vault of liquid metal cooled fast breeder reactors is described. The construction allows relative movement of that part of service cables and pipes which are carried by the fixed roof and that part which is carried by the rotatable plug. (U.K.)

  19. 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.)

  20. Fusion reactor materials

    International Nuclear Information System (INIS)

    At the Belgian Nuclear Research Centre SCK-CEN, activities related to fusion focus on environmental tolerance of opto-electronic components. The objective of this program is to contribute to the knowledge on the behaviour, during and after neutron irradiation, of fusion-reactor materials and components. The main scientific activities for 1997 are summarized

  1. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  2. Cermet fuel reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  3. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  4. 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

  5. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  6. SRP reactor safety evolution

    International Nuclear Information System (INIS)

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

  7. Fusion reactor materials

    International Nuclear Information System (INIS)

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

  8. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1980 are described. The work is presented in three chapters: General Information on the Department, Summary of the Department's Development during 1980, and Activities of the Department. Lists of staff, publications, computer programs, and test facilities are included. (author)

  9. The AP1000 reactor

    International Nuclear Information System (INIS)

    The design of the AP1000 reactor began 20 years ago when Westinghouse launched the AP600 reactor project. In fact by re-assessing AP600's safety margins Westinghouse realized that the its power output could be raised without putting at risk its safety standard. The AP1000 was born, it yields 1100 MWe. The main AP1000's design features is its passive safety (particularly after the Fukushima accident) and its modularity. The passive safety of the AP1000 implies: -) no humane intervention needed for 72 hours at least after the incident; -) no necessity for redundant complex safety systems. The modularity means that the plant, the reactor and other buildings are constructed from a choice of 300 modular units. These units can be built off-site and fit together on site. The modularity allows more construction activities to be led simultaneously and more chances to cope with the construction schedule. The NRC has approved the operation license for 30 years of the first AP1000 being built in the Usa (Vogtle plant in Georgia). 4 AP1000 are being built in China (Sanmen and Haiyang sites) and 6 others are planned in the Usa. Westinghouse is convinced that the AP1000's passive safety makes it more attractive. Let us not forget that Westinghouse was at the origin of the concept of pressurized water reactors, an idea adopted for half the nuclear power stations in the world and for all the plants now active in France. (A.C.)

  10. 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

  11. Pressure tube type reactor

    International Nuclear Information System (INIS)

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  12. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    The general design, characteristics and parameters of TRIGA reactors and fuel are described. This is a training module with the learning objectives: to understand the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and realize the differences between TRIGA fuels and other more traditional. 10 figs., 6 tabs. (nevyjel)

  13. SNAP Nuclear Space Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Corliss, William R

    1966-01-01

    This booklet describes the principles of nuclear-reactor space power plants and shows how they will contribute to the exploration and use of space. It compares them with chemical fuels, solar cells, and systems using energy from radioisotopes. The SNAP (Systems for Nuclear Auxiliary Power) Program, begun in 1955, is described.

  14. 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)

  15. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP)

  16. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  17. Reactor power measuring device

    International Nuclear Information System (INIS)

    The device of the present invention comprises a γ-thermometer disposed in a BWR type reactor, a first amplifier for amplifying the output thereof, a fission ionization chamber disposed in the reactor separately from the γ-thermometer, a second amplifier for amplifying the output thereof, a differential circuit for differentiating the output signal of the second amplifier and a first adding circuit for adding an output signal of the differential circuit and an output signal of the first amplifier. Alternatively, a γ-ray self-powered neutron detector may be disposed instead of the fission ionization chamber. A second adding circuit is also disposed for adding the output signals of plurality of differentiation circuits and inputting the result to the first adding circuit. A sensitivity controller is disposed upstream of the first adder for controlling the sensitivity of the fission ionization chamber. Then, even if time delay should be caused in the γ-thermometer, output signals with good time response characteristic can be obtained by using signals of LPRM or SPND, and currently changing output of the reactor can be measured accurately to provide an effect on the improvement of the safety and operation controllability of the reactor. (N.H.)

  18. 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.)

  19. Reactor physics problems on HCPWR

    International Nuclear Information System (INIS)

    Reactor physics problems on high conversion pressurized water reactors (HCPWRs) are discussed. Described in this report are outline of the HCPWR, expected accuracy for the various reactor physical qualities, and method for K-effective calculation in the resonance energy area. And requested further research problems are shown. The target value of the conversion ratio are also discussed. (author)

  20. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core