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

International Nuclear Information System (INIS)

Wu Xingman

2011-01-01

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)

KAERI results for BN600 full MOX benchmark (Phase 4)

International Nuclear Information System (INIS)

Lee, Kibog Lee

2003-01-01

The purpose of this document is to report the results of KAERI's calculation for the Phase-4 of BN-600 full MOX fueled core benchmark analyses according to the RCM report of IAEA CRP Action on U pdated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. T he BN-600 full MOX core model is based on the specification in the document, F ull MOX Model (Phase4. doc ) . This document addresses the calculational methods employed in the benchmark analyses and benchmark results carried out by KAERI

Planning of the BN-350 reactor decommissioning

International Nuclear Information System (INIS)

Klepikov, A.Kh.; Tazhibayeva, I.L.; Zhantikin, T.M.; Baldov, A.N.; Nazarenko, P.I.; Koltyshev, S.M.; Wells, P.B.

2002-01-01

The experimental and commercial BN-350 NPP equipped with a fast neutron sodium cooled reactor is located in Kazakhstan near the Aktau city on the Caspian Sea coast. It was commissioned in 1973 and intended for weapon-grade plutonium production and as stream supply to a water desalination facility and the turbines of the Mangyshlak Atomic Energy Complex. Taking into account technical, financial and political issues, the Government of Kazakhstan enacted the Decree no. 456 'On Decommissioning of the Reactor BN-350 in the Aktau City of the Mangystau Region'. Because the decision on reactor decommissioning was adopted before the end of scheduled operation (2003), the plan to decommission the BN-350 reactor has not yet been developed. To determine the activities required for ensuring reactor safety and in preparation for decommission in the period prior, the development and ensuring approval by the Republic of Kazakhstan Government of the decommissioning plan, a 'Plan of Priority Actions for BN-350 Reactor Decommissioning' was developed and approved. Actions provided for in the plan include the following: Development of BN-350 Reactor Decommissioning Plan; Accident prevention during the period of transition; Unloading nuclear fuel from reactor and draining the coolant from the heat exchange circuits. Decommission is defined as a complex of administrative and technical actions taken to allow the removal of some or all of regulatory controls over a nuclear facility. These actions involve decontamination, dismantling and removal of radioactive materials, waste, components and structures. They are carried out to achieve a progressive and systematic reduction in radiological hazards and are undertaken on the basis of planning and assessment in order to ensure safety decommissioning operations. In accordance with the decision of Kazakhstan Government, three basic stages for BN-350 reactor decommissioning are envisaged: First stage - Placement of BN-350 into long-term storage

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

International Nuclear Information System (INIS)

Moiseev, A.V.; Khomyakov, Yu.S.; Surov, S.V.

2013-01-01

This paper presents the results of experimental and theoretical work done in 2003-2010 years on substantiation of neutron-physical characteristics of the BN-600 core. 1. Transition to the new core 01M2 with high burnup 11.2% h.a. (the 4-th upgrade of the BN-600 core). Transfer was made without changing the constructive of the core almost by reducing conservatism of design decisions. 2. The end of BN-600 design life cycle and extending it to 10-15 years. Need for analysis and comprehension of the BN-600 experience. 3. Development and introduction of new methods of analysis (precision method of Monte Carlo). 4. In the experiments was a change of equipment and measurement techniques

Results of examinations of safety experimental rods of trap-like type irradiated in the BN-600 reactor

International Nuclear Information System (INIS)

Tarasikov, V.P.; Voznesenskij, R.M.; Rudenko, V.A.

2001-01-01

The results of post-irradiation examination are reported for three trap-like scram rods having been in operation in BN-600 reactor for 311, 331 and 348 EFPD (effective full power days). Compacted boron carbide enriched with 80 at.% 10 B is used as an absorber, zirconium hydride serves as a moderator; cladding are fabricated from steel 06Kh16N15M3B. The results obtained show that two zones are formed in the absorber material which are different in fracture mode and positioned at different distances from the moderator. Radiation damages of steel cladding are noted to be arranged nonuniformly through the height of the rod. The cladding-absorber interaction manifests itself in various ways, this is associated with various absorber burnups. The area of cladding-zirconium hydride interaction constitutes ∼ 30-80 μm. The swelling of the moderator is 4-5% and does not result in a loss of a cladding-moderator clearance [ru

BN-1200 Reactor Power Unit Design Development

International Nuclear Information System (INIS)

Vasilyev, B.A.; Shepelev, S.F.; Ashirmetov, M.R.; Poplavsky, V.M.

2013-01-01

Main goals of BN-1200 design: • Develop a reliable new generation reactor plant for the commercial power unit with fast reactor to implement the first-priority objectives in changing over to closed nuclear fuel cycle; • Improve technical and economic indices of BN reactor power unit to the level of those of Russian VVER of equal power; • Enhance the safety up to the level of the requirements for the 4th generation RP

JNC results of BN-600 benchmark calculation (phase 4)

International Nuclear Information System (INIS)

Ishikawa, Makoto

2003-01-01

The present work is the results of JNC, Japan, for the Phase 4 of the BN-600 core benchmark problem (Hex-Z fully MOX fuelled core model) organized by IAEA. The benchmark specification is based on 1) the RCM report of IAEA CRP on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of LMFR Reactivity Effects, Action 3.12' (Calculations for BN-600 fully fuelled MOX core for subsequent transient analyses). JENDL-3.2 nuclear data library was used for calculating 70 group ABBN-type group constants. Cell models for fuel assembly and control rod calculations were applied: homogeneous and heterogeneous (cylindrical supercell) model. Basic diffusion calculation was three-dimensional Hex-Z model, 18 group (Citation code). Transport calculations were 18 group, three-dimensional (NSHEC code) based on Sn-transport nodal method developed at JNC. The generated thermal power per fission was based on Sher's data corrected on the basis of ENDF/B-IV data library. Calculation results are presented in Tables for intercomparison

Neutronic safety parameters of the BN-600 type reactor with hybrid core. Diffusion and transport approach. R-Z homogeneous media

International Nuclear Information System (INIS)

Cherny, V.; Danilytchev, A.; Korobeinikov, V.; Korobeinikova, L.; Stogov, V.

2000-01-01

The present paper includes the results of neutronic safety calculations of the BN-600 hybrid core benchmark problem. Results presented include: multiplication factors, Doppler coefficients, fuel and structure density coefficients, expansion coefficients, power distribution, beta-effective values, reaction rate distributions

Safety Design Criteria and Approaches to Safety Substantiation of the BN-1200

International Nuclear Information System (INIS)

Ashurko, I.

2013-01-01

Russian experience in SFR area: Activities on development of safety design criteria for SFRs of the 4th generation is carried out within the GIF framework. Although this reactor technology is considered as innovative that is relevant to the 4th generation, however, it has already a certain history. In this relation, it seems to be useful to analyze the corresponding experience that is available in various countries. 4 SFRs have been successfully operated in the USSR and in the Russian Federation: • Experimental reactor BR-5/10; • Research reactor BOR-60; • Prototype BN-350 power reactor; • Commercial BN-600 power unit at the Beloyarsk NPP. Thus, Russia gained a considerable experience of design, construction and operation of SFRs. In particular, a certain experience has been acquired on safety substantiation of reactors of this type and their licensing. Now BOR-60 and BN-600 continue their operation, BN-800 power unit is under construction, development of the commercial BN-1200 power unit, that is considered as the 4th generation reactor, has been started. Due to limited number of operating SFRs in the world, successful Russian experience in this area should be taken into account for further development and improvement of SFR SDC developed by the GIF Task Force. In particular, participation of SFR designers in this activities would be fruitful and useful

Status of fast reactor activities in the USSR

International Nuclear Information System (INIS)

Troyanov, M.F.; Rinejskij, A.A.

1990-01-01

Four fast reactors are in operation in the USSR now: BR-10, BOR-60, BN-350 and BN-600. Load factor of BN-600 reactor was in 1989 about 76%. On the basis of operational experience of running reactors design of more powerful commercial size BN-800 power reactor has been completed recently and construction work has started at two sites. The BN-1600 reactor is considered to be the prototype of future commercial reactors. In 1990, it was decided to extend its design approach with the aim to find some additional solutions to provide higher safety and better economics. (author). Figs and tabs

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

International Nuclear Information System (INIS)

Grachyov, A.F.; Kalygin, V.V.; Skiba, O.V.; Mayorshin, A. A.; Bychkov, A.V.; Kisly, V.A.; Ovsyannikov, Y.F.; Bobrov, D.A.; Mamontov, S.I.; Tsyganov, A.N.; Churutkin, E.I.; Davydov, P.I.; Samosenko, E.A; Shalak, A.R.; Ojima, Hisao

2004-01-01

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

An analysis of main processes at small water-into-sodium leaks in the BN-350 and BN-600 NPP steam generators

International Nuclear Information System (INIS)

Poplavsky, V.M.

1990-01-01

The paper presents the main characteristics of emergency processes at small water-into-sodium leaks that took place during the BN-350 and BN-600 NPP steam generators operation. Leak characteristics are presented, the relationship between such parameters as leak rate and duration, its location in a tube bundle, mass of water ingress into sodium, and the character and size of a failure in the interaction zone is analyzed. (author). 5 refs, 3 figs, 2 tabs

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

International Nuclear Information System (INIS)

Ashurko, Y.M.; Baklushin, R.P.; Zagorulko, Y.I.; Ivanenko, V.N.; Matveyev, V.P.; Vasilyev, B.A.

2000-01-01

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)

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

International Nuclear Information System (INIS)

Ishikawa, M.

2002-01-01

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

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)

2013-12-01

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

Status of fast reactor activities in the Russian Federation

Energy Technology Data Exchange (ETDEWEB)

Troyanov, M F; Rinejsjij, A A [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

1992-07-01

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.

Status of fast reactor activities in the Russian Federation

International Nuclear Information System (INIS)

Troyanov, M.F.; Rinejsjij, A.A.

1992-01-01

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

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)

BN-800 as a new stage in development of fast neutron sodium cooled reactors

International Nuclear Information System (INIS)

Poplavskij, V.M.; Chebeskov, A.N.; Matveev, V.I.

2004-01-01

The role of fast reactors in the strategy of evolution of the nuclear power of Russia is discussed, BN-800 under construction, where unique technical and construction decisions are used, is viewed. Economical estimations of expenses with regard for all life cycle demonstrate that fast reactors may be no higher-priced than the most popular in the world water moderated reactors. Closing of nuclear fuel cycle of BN-800 makes possible decision of the problem of plutonium and actinide utilization, that makes the fast reactor more safety for the environment [ru

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

International Nuclear Information System (INIS)

Poplavski, V.M.; Ashurko, Y.M.; Zverev, K.V.; Sarayev, O.M.; Oshkanov, N.N.; Korol'kov, A.S.

1999-01-01

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)

Status of fast reactor activities in Russia

International Nuclear Information System (INIS)

Poplavski, V.M.; Ashurko, Yu.M.; Zverev, K.V.

1998-01-01

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)

Fast reactors in nuclear power

Energy Technology Data Exchange (ETDEWEB)

Kazachkovskii, O

1981-02-01

The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

Fast reactors in Russia: State of the art and trends of development

International Nuclear Information System (INIS)

Poplavsky, V.M.; Ashurko, Yu.M.; Zverev, K.V.; Oshkanov, N.N.; Korol'kov, A.S.; Filin, A.I.

2002-01-01

This status report contains the following: facts on nuclear power in Russia from 2001-2002; plans for further development of nuclear power; state of the art on operation of fast reactors in 2002, namely BN-600, experimental reactors BOR-60 and BR-10; construction of NPP BN-800; participation in activities on BN-350 reactor decommissioning; description of trends of design studies in the field of fast reactors and accelerator driven systems

Immobilization of Cesium Traps from the BN-350 Fast Reactor (Aktau, Kazakhstan)

Energy Technology Data Exchange (ETDEWEB)

J. A. Michelbacher; C. Knight; O. G. Romanenko; I. L. Tazhibaeva; I. L. Yakovlev; A. V. Rovneyko; V. I. Maev; D. Wells; A. Herrick

2011-03-01

During BN-350 reactor operations and also during the initial stages of decommissioning, cesium traps were used to decontaminate the reactor’s primary sodium coolant. Two different types of carbon-based trap were used – the MAVR series, low ash granulated graphite adsorber (LAG) contained in a carrier designed to be inserted into the reactor core during shutdown; and a series of ex-reactor trap accumulators(TAs) which used reticulated vitreous carbon (RVC) to reduce Cs-137 levels in the sodium after final reactor shutdown. In total four MAVRs and seven TAs were used at BN-350 to remove an estimated cumulative 755 TBq of cesium. The traps, which also contain residual sodium, need to be immobilized in an appropriate way to allow them to be consigned as waste packages for long term storage and, ultimately, disposal. The present paper reports on the current status of the implementation phase, with particular reference to the work done to date on the trap accumulators, which have the most similarity with the cesium traps used at other reactors.

Cobalt-60 production in the BN-350 fast power reactor

International Nuclear Information System (INIS)

Zvonarev, A.V.; Korobejnikov, V.V.; Matveenko, I.P.

1994-01-01

A possibility of Co-60 isotope production in the BN-350 fast reactor was considered. A special irradiating device, which is an assembly with a central hole, where a container containing cobalt and zirconium hydride is placed. The irradiating device tested permits generating 60 Co with specific activity of 100 Ci/g

Chemical surveillance of commercial fast breeder reactors

International Nuclear Information System (INIS)

Stamm, H.H.; Stade, K.Ch.

1988-01-01

After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

k-eff of the Bn-350 reactor fuel by transportation

International Nuclear Information System (INIS)

Lado, A. V.; Romanenko, O. G.; Tazhibaeva, I. L.

2001-01-01

There is packaging of nuclear fuel on the BN-350 fast breeder reactor, Actau, now. The analysis of criticality while this procedure was done in the Safety Analysis Report . Keeping in mind the planning displacement of the fuel to a site of long-term storage, the criticality assessment of the fuel packed into transportation cask carried out in this paper

UK contributions to the decommissioning of the BN-350 reactor in Kazakhstan: 2002 – 2011

International Nuclear Information System (INIS)

Wells, D.

2011-01-01

UK assistance with the decommissioning of BN-350 has cost ~£8.9 million over ten years, ~£4 million spent directly in Kazakhstan. The Programme has immobilised key wastes, contributed to irreversible shutdown of the reactor and addressed issues associated with sodium coolant processing. The Programme funded the operations to load spent fuel canisters into casks at BN-350, together with their despatch from site and receipt at the secure storage facility. The Programme also delivered technical and project management training, assisted in the production of the BN-350 Decommissioning Plan and contributed to the radiation survey effort in the STS

An experience of cleaning and decontamination of the BN-350 reactor components

International Nuclear Information System (INIS)

Vasilenko, K.T.; Kochetkov, L.A.; Arkhipov, V.M.; Baklushin, R.P.; Gorlov, A.I.; Kiselev, G.V.; Rezinkin, P.S.; Samarkin, A.A.; Tverdovsky, N.D.

1978-01-01

In the course of start-up, adjustment and operation of the BN-350 reactor there arose a need for cleaning from sodium and decontamination of primary and secondary equipment components. Design schemes of the systems provided for this purpose as well as those specially designed for cleaning of steam generator evaporators are considered. Technological processes of cleaning and decontamination for some reactor components (removable parts of circulating pumps, evaporators, valves) are described, the results are presented. (author)

Fuel component of electricity generation cost for the BN-800 reactor with MOX fuel and uranium oxide fuel with increasing of fuel burnup and removing of radial breeding blanket

International Nuclear Information System (INIS)

Raskach, A.

2001-01-01

Nowadays there are two completed design concepts of Nuclear Power Plants (NPPs) with the BN-800 type reactors developed with due regard for advanced safety requirements. One of them is the design of the fourth unit of the Beloyarsk Nuclear Power Plant; the other one is the design of three units of the South Ural Nuclear Power Plant. The both concepts are to use mixed oxide fuel (MOX fuel) based on civil plutonium. Studies on any project include economical analyses and cost of fuel is an essential parameter. In the course of the design works on the both projects such evaluations were done. For BN-800 on the Beloyarsk site nuclear fuel costs were taken from actual expenses of the BN-600 reactor and converted to rated thermal power and design capacity factor of the BN-800 and then increased by 20% in connection with turning to MOX fuel. Then this methodology was rewarding, but the ratio of uranium fuel and MOX fuel costs might change for the last years. For the project of three units of the South Ural Nuclear Power Plant nuclear fuel expenses were calculated from the data on a MOX fuel fabrication production facility (Complex-300). However, investigations performed recently shown that the methodology of economical assessments should be revised, as well as design and technology of MOX fuel fabrication at Complex-300 should be revised to meet all the existing safety requirements. Excepting there is a great bulk of civil plutonium to be reproduced, now we came up against the problem to utilize the exceeding ex-weapons plutonium that obviously can be used for MOX fuel fabrication as well. Construction of the MOX fuel fabrication facility - Complex-300 - was started in 1983. Its design output was planned to provide simultaneously 4 fast reactors of the BN-800 type with MOX fuel. By now about 50% of construction works (taking into account auxiliary buildings and arrangements) and 20% of installation works have been done at Complex-300. Along this, first works to construct

Preliminary design of reactor coolant pump canned motor for AC600

International Nuclear Information System (INIS)

Deng Shaowen

1998-01-01

The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

BN-350 nuclear power plant. Regulatory aspects of decommissioning

International Nuclear Information System (INIS)

Shiganakov, S.; Zhantikin, T.; Kim, A.

2002-01-01

Full text: The BN-350 reactor is a fast breeder reactor using liquid sodium as a coolant [1]. This reactor was commissioned in 1973 and operated for its design life of 20 years. Thereafter, it was operated on the basis of annual licenses, and the final shutdown was initially planned in 2003. In 1999, however, the Government of the Republic of Kazakhstan adopted Decree on the Decommissioning of BN-350 Reactor. This Decree establishes the conception of the reactor plant decommissioning. The conception envisages three stages of decommissioning. The first stage of decommissioning aims at putting the installation into a state of long term safe enclosure. The main goal is an achievement of nuclear-and radiation-safe condition and industrial safety level. The completion criteria for the stage are as follows: spent fuel is removed and placed in long term storage; radioactive liquid metal coolant is drained from the reactor and processed; liquid and solid radioactive wastes are reprocessed and long-term stored; systems and equipment, that are decommissioned at the moment of reactor safe store, are disassembled; radiation monitoring of the reactor building and environment is provided. The completion criteria of the second stage are as follows: 50 years is up; a decision about beginning of works by realization of dismantling and burial design is accepted. The goal of the third stage is partial or total dismantling of equipment, buildings and structure and burial. Since the decision on the decommissioning of BN-350 Reactor Facility was accepted before end of scheduled service life (2003), to this moment 'The Decommissioning Plan' (which in Kazakhstan is called 'Design of BN-350 reactor Decommission') was not worked out. For realization of the Governmental Decree and for determination of activities by the reactor safety provision and for preparation of its decommission for the period till Design approval the following documents were developed: 1. Special Technical Requirements

Overview of the fast reactors fuels program

International Nuclear Information System (INIS)

Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

1980-04-01

Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

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.

Assessment of the dry process fuel sodium-cooled fast reactors

International Nuclear Information System (INIS)

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 ∼50% and most of the fission products were removed

Accidental situations analysis in BN-800 reactor bounded with an untimely ascension of the control rods

International Nuclear Information System (INIS)

Gorbunov, V.S.; Zaets, N.P.

1987-12-01

In this document the conditions and the results of one or more control rods untimely ascension out of the BN-800 core are examined. The mathematical model which describes the reactor kinetic, the temperature core and the feedback is presented [fr

Overview of the fast reactors fuels program. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

1980-04-01

Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

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

International Nuclear Information System (INIS)

Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

2012-09-01

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

Concept and designs of new-generation fast reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.

1993-01-01

This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

Comparison of fuel cycles characteristics for nuclear energy systems based on WWER-TOI and BN-1200 reactors

International Nuclear Information System (INIS)

Kagramanyan, V.S.; Kalashnikov, A.G.; Kapranova, Eh.N.; Puzakov, A.Yu.

2014-01-01

Authors determine the characteristics of the fuel cycle (FC) based on stationary nuclear power system based on WWER-TOI and BN-1200 reactors with fuel of different composition. Characteristics of reactor systems with partial or complete spent nuclear fuel reprocessing and recycling of plutonium are compared to those of the reference system consisting only of WWER-TOI with uranium oxide fuel, operating in an open FC [ru

Structural Integrity Evaluation of the KALIMER-600 Reactor Core Support Structure

International Nuclear Information System (INIS)

Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

2005-01-01

KALIMER-600(Korea Advanced LIquid MEtal Reactor, 600MWe) is a pool type sodium-cooled liquid metal reactor. Since the normal operating temperature of KALIMER-600 is 545 .deg. C, the reactor structures in the hot pool region are designed and evaluated according to the elevated temperature design rules such as the ASME Boiler and Pressure Vessel Code Section III, Subsection NH. Since the core support structure of KALIMER-600 is in the cold pool region under 400 .deg. C, a high temperature inelastic behavior is not expected. Thus the stress and fatigue limits are the main concerns to assure the structural design integrity following the ASME Subsection NG. In this paper, the evaluations of the stress and fatigue damage for the core support structure of KALIMER-600 are carrried out in the case of a normal operation condition using the rules of ASME Subsection NG. To obtain the stress values, a heat transfer analysis and a stress analysis under a combined loading condition are performed. From the stress distribution results, the critical sections are selected and the stress and fatigue limits are evaluated for the selected regions

Applicability of RELAP5 for safety analysis of AP600 and PIUS reactors

International Nuclear Information System (INIS)

Motloch, C.G.; Modro, S.M.

1990-01-01

An assessment of the applicability of using RELAP5 for performing safety analyses of the AP600 and PIUS advanced reactor concepts is being performed. This ongoing work is part of a larger safety assessment of advanced reactors sponsored by the United States Nuclear Regulatory Commission. RELAP5 models and correlations are being reviewed from the perspective of the new AP600 and PIUS phenomena and features that could be important to reactor safety. The purpose is to identify those areas in which new mathematical models of physical phenomena would be required to be added to RELAP5. In most cases, the AP600 and PIUS designs and systems and the planned and off-normal operations are similar enough to current Pressurized Water Reactors (PWR) that RELAP5 safety analysis applicability is unchanged. However, for AP600 the single most important systemic and phenomenological difference between it and current PWRs is in the close coupling between the reactor system and the containment during postulated Loss of Coolant Accident (LOCA) events. This close coupling may require the addition of some thermal-hydraulic models to RELAP5. And for PIUS, the most important new feature is the thermal density locks. These and other important safety-related features are discussed. This document presents general descriptions of RELAP5, AP600, and PIUS, describes the new features and phenomena of the reactors, and discusses the code/reactors safety-related issues. 32 refs., 4 figs., 2 tabs

Experience on Russian military origin plutonium conversion into fast reactor nuclear fuel

International Nuclear Information System (INIS)

Grachev, A.F.; Skiba, O.V.; Bychkov, A.V.; Mayorshin, A.A.; Kisly, V.A.; Bobrov, D.A.; Osipenko, A.G.; Babikov, L.G.; Mishinev, V.B.

2001-01-01

According to the Concept of Russian Minatom on military plutonium excess utilization, the State Scientific Center of Russian Federation ''Research Institute of Atomic Reactors'' (Dimitrovgrad) has begun study on possibility of technological processing of the metal military plutonium into MOX fuel. The Program and the stages of its realization are submitted in the paper. During 1998-2000 the first stage of the Program was fulfilled and 50 kg of military origin metallic plutonium was converted to MOX fuel for the BOR-60 and BN-600 reactor. The plutonium conversion into MOX fuel is carried out under the original technology developed by SSC RIAR. It includes pyro-electrochemical process for production of fuel on the domestic equipment with the subsequent fuel pins manufacturing for the fast reactors by the vibro-packing method. The produced MOX fuel is purified from alloy additives (Ga) and corresponds to the vibro-packed fuel standard for fast reactors. The fuel pins manufacturing for BOR-60 and BN-600 reactors are carried out by the vibro-packing method on a standard procedure, which is used in SSC RIAR more than 20 years. (author)

Low-pressure c-BN deposition - is a CVD process possible?

International Nuclear Information System (INIS)

Haubner, R.; Tang, X.

2001-01-01

Since the low-pressure diamond deposition was discovered in 1982 there is a high interest to find a similar process for the c-BN synthesis. A review about the c-BN deposition process as well as its characterization is given. Experiments with a simple chemical vapor deposition(CVD) reactor using tris(dimethylamino)borane as precursor were carried out. In a cold-wall reactor substrates were heated up by high-frequency. Argon was used as protecting and carrying the precursor, it was saturated with tris(dimethylamino)borane (precursor) according to its vapor pressure and transports the pressure to the hot substrate, where deposition occurs. WC-Co hardmetal plates containing 6 wt. % Co, Mo and Si were used as substrates. Various BN layers were deposited and characterized. X-ray diffraction, IR-spectroscopy and SIMS indicate that BN-coatings containing c-BN were deposited. However a final verification of c-BN crystallites by TEM investigations was not possible till now. (nevyjel)

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)

Trufanov, A.A.; Orlov, Yu.I.; Sorokin, A.P.; Chernonog, V.L.

2015-01-01

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 [ru

Passive safety design characteristics of the KALIMER-600 burner reactor

International Nuclear Information System (INIS)

Kwon, Young-Min; Jeong, Hae-Yong; Cho, Chung-Ho; Ha, Ki-Seok; Kim, Sang-Ji

2009-01-01

The Korea Atomic Energy Research Institute (KAERI) has recently studied several burner core designs for a transuranics (TRU) transmutation based on the breakeven core geometry of KALIMER-600. The KALIMER-600 is a net electrical rating of 600MWe, sodium-cooled, metallic-fueled, pool-type reactor. For the burner core concept selected for the present analysis, the smearing fractions of the fuel rods in three fuel zones are changed while maintaining the cladding outer diameter and cladding thickness. The resulting fuel slug smearing fractions of the inner, middle, and outer core zones are 36%, 40%, and 48%, respectively. The TRU conversion ratio is 0.57 and the TRU enrichment of the driver fuel is set to 30.0 w/o because of the current practical limitation of the U-TRU-10%Zr metal fuel database. The purpose of this paper is to evaluate the safety performance characteristics provided by the passive safety design features in the KALIMER-600 burner reactor by using a system-wide safety analysis code. The present scoping analysis focuses on an assessment of the enhanced safety design features that provide passive and self-regulating responses to transient conditions and an evaluation of the safety margin during unprotected overpower, unprotected loss of flow, and unprotected loss of heat sink events. The analysis results show that the KALIMER-600 burner reactor provides larger safety margins with respect to the sodium boiling, fuel rod integrity, and structural integrity. The overall inherent safety can be enhanced by accounting for the reactivity feedback mechanisms in the design process. (author)

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

International Nuclear Information System (INIS)

Pechenkin, V.A.

2012-01-01

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

AC-600 reactor reloading pattern optimization by using genetic algorithms

International Nuclear Information System (INIS)

Wu Hongchun; Xie Zhongsheng; Yao Dong; Li Dongsheng; Zhang Zongyao

2000-01-01

The use of genetic algorithms to optimize reloading pattern of the nuclear power plant reactor is proposed. And a new encoding and translating method is given. Optimization results of minimizing core power peak and maximizing cycle length for both low-leakage and out-in loading pattern of AC-600 reactor are obtained

IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

International Nuclear Information System (INIS)

Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

2002-01-01

Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

BN-350 decommissioning problems of radioactive waste management

International Nuclear Information System (INIS)

Galkin, A.; Tkachenko, V.

2002-01-01

Pursuant of modern concept on radioactive waste management applied in IAEA Member States all radioactive wastes produced during the BN-350 operation and decommissioning are subject to processing in order to be transformed to a form suitable for long-term storage and final disposal. The first two priority objectives for BN-350 reactor are as follows: cesium cleaning from sodium followed by sodium drain, and processing; processing of liquid and solid radioactive waste accumulated during BN-350 operation. Cesium cleaning from sodium and sodium processing to NaOH will be implemented under USA engineering and financial support. However the outputted product might be only subject to temporary storage under special conditions. Currently the problem is being solved on selection of technology for sodium hydroxide conversion to final product incorporated into cement-like matrix ready for disposal pursuant to existing regulatory requirements. Industrial installation is being designed for liquid radioactive waste processing followed by incorporation to cement matrix subject to further disposal. The next general objective is management of radioactive waste expected from BN-350 decommissioning procedure. Complex of engineering-radiation investigation that is being conducted at BN-350 site will provide estimation of solid and liquid radioactive waste that will be produced during the course of the BN-350 decommission. Radioactive wastes that will be produced may be shared for primary (metal structures of both reactor and reactor plant main and auxiliary systems equipment as well as construction wastes of dismantled biological protection, buildings and structures) and secondary (deactivation solutions, tools, materials, cloth, special accessory, etc.). Processing of produced radioactive wastes (including high activity waste) requires the use of special industrial facilities and construction of special buildings and structures for arrangement of facilities mentioned as well as for

A new concept of fast reactors, the potentialities of burning in them of actinoid and weapon-grade plutonium

International Nuclear Information System (INIS)

Murogov, V.M.; Troyanov, M.F.; Ilyunin, V.G.; Rudneva, V.Ya.

1992-01-01

The approach to a possible solution of the problem of peaceful utilization of weapon-grade plutonium released in the result of nuclear disarmament in Russia is given in the repot. As the most safe, ecologically acceptable and economically effective way of the plutonium utilization is the usage of such plutonium as a fuel for atomic power station. It is proposed to decide the problem on the basis of BN-600 and BN-800 reactors. In the approach, thorium could be used as a fertile material. The secondary nuclear fuel U-233 is expedient to use in light-water reactors of new generation. (author)

Nuclear power in Kazakhstan and current status of the BN-350 fast reactor

International Nuclear Information System (INIS)

Krechetov, S.

1998-01-01

Atomic scientific-industrial complex of Republic of Kazakhstan consist of: Number of natural uranium mines, two plants of U 3 O 8 production at Aktau and Stepnogorsk towns, metallurgical plant producing fuel pellets for RBMK and WWER fuel assemblies. Fast breeder reactor with sodium coolant BN - 350 at Aktau. The average share of BN-350 in total electricity production is 0.7%. Taking into account common condition industrial in Kazakhstan have no significant improvement the total electricity production on goal and oil station stayed on the same level as in 1996. According to government decision in 1998 the following structure of atomic complex have been established. Several rather serious events should be mentioned. In January 1998 the Provision of licensing in nuclear field was signed by Prime Ministry and now Kazakhstan have all necessary acts for starting this process. In April 1998 the General Program of development atomic scientific and industrial complex of Kazakhstan had been reported to Government and got approval in whole. In particular this program are including the design and construction NPP for electricity production on the lake Balhash, and two NPP for heating Almaty and new capital Akmola. In April 1998 the law on Radiation protection had got approval of Parliament and now President should sign it. In January the Nuclear Technologies Safety Center (NTSC) had been established by group of organizations such as KAEA, NNC, University, Nuclear Society of Kazakhstan, Center of standardization and Almaty local administration. NTSC have established as a society independent experts in the field nuclear safety. With cooperation with ANL an expertise on nuclear safety of BN-350 will be done related to long-term spent fuel storage

Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

International Nuclear Information System (INIS)

Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

2007-01-01

Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

Analytical and Experimental Study for Validation of the Device to Confine BN Reactor Melted Fuel

International Nuclear Information System (INIS)

Rogozhkin, S.; Osipov, S.; Sobolev, V.; Shepelev, S.; Kozhaev, A.; Mavrin, M.; Ryabov, A.

2013-01-01

To validate the design and confirm the design characteristics of the special retaining device (core catcher) used for protection of BN reactor vessel in the case of a severe beyond-design basis accident with core melting, computational and experimental studies were carried out. The Tray test facility that uses water as coolant was developed and fabricated by OKBM; experimental studies were performed. To verify the methodical approach used for the computational study, experimental results obtained in the Tray test facility were compared with numerical simulation results obtained by the STAR-CCM+ CFD code

Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

International Nuclear Information System (INIS)

Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum

2005-01-01

KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types

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.

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

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2005-09-01

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 ∼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

Review of the proposed materials of construction for the SBWR and AP600 advanced reactors

International Nuclear Information System (INIS)

Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F.

1994-06-01

Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited

An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

Energy Technology Data Exchange (ETDEWEB)

Park, Chang Gyu; Lee, Jae Han

2007-03-15

This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

Direct conversion of h-BN into c-BN and formation of epitaxial c-BN/diamond heterostructures

International Nuclear Information System (INIS)

Narayan, Jagdish; Bhaumik, Anagh; Xu, Weizong

2016-01-01

We have created a new state of BN (named Q-BN) through rapid melting and super undercooling and quenching by using nanosecond laser pulses. Phase pure c-BN is formed either by direct quenching of super undercooled liquid or by nucleation and growth from Q-BN. Thus, a direct conversion of hexagonal boron nitride (h-BN) into phase-pure cubic boron nitride (c-BN) is achieved by nanosecond pulsed laser melting at ambient temperatures and atmospheric pressure in air. According to the P-T phase diagram, the transformation from h-BN into c-BN under equilibrium processing can occur only at high temperatures and pressures, as the hBN-cBN-Liquid triple point is at 3500 K/9.5 GPa or 3700 K/7.0 GPa with a recent theoretical refinement. Using nonequilibrium nanosecond laser melting, we have created super undercooled state and shifted this triple point to as low as 2800 K and atmospheric pressure. The rapid quenching from super undercooled state leads to the formation of a new phase, named as Q-BN. We present detailed characterization of Q-BN and c-BN layers by using Raman spectroscopy, high-resolution scanning electron microscopy, electron-back-scatter diffraction, high-resolution TEM, and electron energy loss spectroscopy, and discuss the mechanism of formation of nanodots, nanoneedles, microneedles, and single-crystal c-BN on sapphire substrate. We have also deposited diamond by pulsed laser deposition of carbon on c-BN and created c-BN/diamond heterostructures, where c-BN acts as a template for epitaxial diamond growth. We discuss the mechanism of epitaxial c-BN and diamond growth on lattice matching c-BN template under pulsed laser evaporation of amorphous carbon, and the impact of this discovery on a variety of applications.

Ideas in support to the definition of the Phase 6

International Nuclear Information System (INIS)

Rimpault, G.

2004-01-01

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

The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

CERN Document Server

Lestone, J P; Rennie, J A; Sprinkle, J K; Staples, P; Grimm, K N; Hill, R N; Cherradi, I; Islam, N; Koulikov, J; Starovich, Z

2002-01-01

The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as approx 10 sup 5 Rads/h. Using limited facility information and multiple measurements along the length of spe...

Under-Sodium Inspection Techniques for Reactor Internals of KALIMER-600 using Ultrasonic Waveguide Sensor

International Nuclear Information System (INIS)

Joo, Young Sang; Kim, Seok Hoon; Lee, Jae Han

2005-01-01

KALIMER-600 is a pool type liquid metal reactor (LMR) which is operated with a sodium coolant. The reactor internals of KALIMER-600 are submerged in a liquid sodium pool. As the liquid sodium is opaque to the light, a conventional visual inspection can not be used for observing the internal structures under a sodium condition. An under-sodium viewing (USV) technique using an ultrasonic wave should be applied for the observation of the refueling maneuver and the in-service inspection of the reactor internals. Under-sodium inspection technology utilizing ultrasonic waves has been widely developed for a visualization of the reactor core and internal components of LMR. Immersion sensors and waveguide sensors have been applied to the USV inspection. The immersion sensor has a precise imaging capability, but may have high temperature restrictions and an uncertain life. The waveguide sensor has the advantages of simplicity and reliability, but limited in its movement. The new plate-type waveguide sensor has been developed as a useful alternative to immersion sensors for USV applications. In the viewing and monitoring applications, a beam steering function of a waveguide sensor might be required. A new waveguide sensor and technique are being developed to overcome the limitations of a waveguide ultrasonic sensor. In this study, the under-sodium inspection techniques using the newly developed waveguide sensor for the reactor internal structures of KALIMER-600 is proposed

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)

Buksha, Yu.K.; Zabudko, L.M.; Kravchenko, I.N.; Matveenko, L.V.; Meshkov, M.N.

1984-01-01

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)

The KALIMER-600 Reactor Core Design Concept with Varying Fuel Cladding Thickness

International Nuclear Information System (INIS)

Hong, Ser Gi; Jang, Jin Wook; Kim, Yeong Il

2006-01-01

Recently, Korea Atomic Energy Research Institute (KAERI) has developed a 600MWe sodium cooled fast reactor, the KALIMER-600 reactor core concept using single enrichment fuel. This reactor core concept is characterized by the following design targets : 1) Breakeven breeding (or fissile-self-sufficient) without any blanket, 2) Small burnup reactivity swing ( 23 n/cm 2 ). In the previous design, the single enrichment fuel concept was achieved by using the special fuel assembly designs where non-fuel rods (i.e., ZrH 1.8 , B 4 C, and dummy rods) were used. In particular, the moderator rods (ZrH 1.8 ) were used to reduce the sodium void worth and the fuel Doppler coefficient. But it has been known that this hydride moderator possesses relatively poor irradiation behavior at high temperature. In this paper, a new core design concept for use of single enrichment fuel is described. In this concept, the power flattening is achieved by using the core region wise cladding thicknesses but all non-fuel rods are removed to simplify the fuel assembly design

Multi-Functional All BN-BN Composites

Data.gov (United States)

National Aeronautics and Space Administration — Development of multifunctional Boron Nitride nanotube-Boron Nitride (BN-BN) composites to provide novel energy transducers, thermal conductors, anti-penetrator/wear...

Nuclear reactor fuel cycle technology with pyroelectrochemical processes

International Nuclear Information System (INIS)

Skiba, O.V.; Maershin, A.A.; Bychkov, A.V.; Zhdanov, A.N.; Kislyj, V.A.; Vavilov, S.K.; Babikov, L.G.

1999-01-01

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 [ru

Microstructural investigations of fast reactor irradiated austenitic and ferritic-martensitic stainless steel fuel cladding

International Nuclear Information System (INIS)

Agueev, V.S.; Medvedeva, E.A.; Mitrofanova, N.M.; Romanueev, V.V.; Tselishev, A.V.

1992-01-01

Electron microscopy has been used to characterize the microstructural changes induced in advanced fast reactor fuel claddings fabricated from Cr16Ni15Mo3NbB and Cr16Ni15Mo2Mn2TiVB austenitic stainless steels in the cold worked condition and Cr13Mo2NbVB ferritic -martensitic steel following irradiation in the BOR-60, BN-350 and BN-600 fast reactors. The data are compared with the results obtained from a typical austenitic commercial cladding material, Cr16Ni15Mo3Nb, in the cold worked condition. The results reveal a beneficial effect of boron and other alloying elements in reducing void swelling in 16Cr-15Ni type austenitic steels. The high resistance of ferritic-martensitic steels to void swelling has been confirmed in the Cr13Mo2NbVB steel. (author)

Current status of work on preservation of accumulated knowledge on fast reactors in Russia and plan of top-priority measures

International Nuclear Information System (INIS)

Kotchetkov, L.A.; Poplavsky, V.M.; Tsiboulya, A.M.; Ashurko, Yu.M.

2005-01-01

The future of nuclear power is associated with mastering of fast reactor technology. Experience gained in Russia by now in the development of sodium cooled fast reactors (FR) and reactor plants of various applications is one of the most extensive and successful all over the world. Since the late 1940-ies up to now, well-directed, rather intensive work has been carried out in the USSR (later in Russia) on all aspects of fast reactors. Institute for Physics and Power Engineering has been always leading organization in the USSR and in Russia in the area of fast reactors. Work on fast reactors in Russia was carried out by many institutions, namely: IPPE, VNIINM, OKBM, VNIPIET, OKB Gidropress, RIAR, SPbAEP, TsNII Prometey and teams of the BN-350 and the BN-600 plants working in close and fruitful cooperation. Successful operation of the power unit of Beloyarskaya NPP with the BN-600 fast reactor during 25 years is one of the good results of this vast expensive efforts. In view of delay in wide-scale deployment of fast neutron reactors and change of generations of specialists in the area of FR, a necessity has arisen in the preservation of knowledge and experience on FR gained in many countries including Russia. Certain measures in this area have been planned by the Russian organizations. However, the necessity has become imminent in a purposeful systematized approach to the preservation of knowledge in fast reactor area, which can be realized only within the framework of development of appropriate special work program. The basic work trends within the framework of this program have been stated. In view of urgency of some part of this work, it is necessary to prioritize the work contents. IAEA assistance (methodological, organizational and financial) in the implementation of some sections of the program would facilitate successful implementation of the work program on preservation of knowledge on FR in Russia. (author)

Few-group constants for the calculation of ksub(eff) and Δ(1/ksub(eff)) of fast breeder reactors

International Nuclear Information System (INIS)

Svarny, J.

1978-01-01

A theoretical and numerical analysis is presented of the linear and bilinear weighting of group constants. Special attention is paid to error accumulation in the few-group calculations of reactivity (ksub(eff)) and its first order perturbations caused by inaccuracies in weighting functions. Some theoretical conclusions are supported by calculations of the BN-600 fast breeder reactor. (author)

Fuel component of electricity generation cost for the BN-800 reactor with 800 MOX fuel and uranium oxide fuel, increased fuel burnup, and removal of radial breeding blanket

International Nuclear Information System (INIS)

Raskach, A.

2000-01-01

There are two completed design concepts of NPP with BN-800 type reactors developed with due regard for enhanced safety requirements. They have been created for the 3 rd unit of Beloyarsk NPP and for three units of South Ural NPP. Both concepts are proposed to use mixed oxide fuel (MOX) based on civil plutonium. At this moment economical estimations carried out for these projects need to be revised in connection with the changes of economical situation in Russia and the world nuclear market structure. It is also essential to take into account the existing problem of the excess ex-weapons plutonium utilization and the possibility of using this plutonium to fabricate MOX fuel for the BN-800 reactors. (authors)

Calculation of criticality of the AP600 reactor with KENO V.a code

Energy Technology Data Exchange (ETDEWEB)

Krumbein, A; Caner, M; Shapira, M [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center

1996-12-01

The Westinghouse AP600 PWR has been modeled using the KENO V.a three dimensional Monte Carlo criticality program of the SCALE-PC code system. These calculations and the use of a Monte Carlo neutron transport code such as KENO will provide us with an independent check on our WIMS/CITATION calculations for the AP600 as well as for other reactors. It will also enable us to model more complicated geometries. (authors).

Pressure sensing element based on the BN-graphene-BN heterostructure

Science.gov (United States)

Li, Mengwei; Wu, Chenggen; Zhao, Shiliang; Deng, Tao; Wang, Junqiang; Liu, Zewen; Wang, Li; Wang, Gao

2018-04-01

In this letter, we report a pressure sensing element based on the graphene-boron nitride (BN) heterostructure. The heterostructure consists of monolayer graphene sandwiched between two layers of vertically stacked dielectric BN nanofilms. The BN layers were used to protect the graphene layer from oxidation and pollution. Pressure tests were performed to investigate the characteristics of the BN-graphene-BN pressure sensing element. A sensitivity of 24.85 μV/V/mmHg is achieved in the pressure range of 130-180 kPa. After exposing the BN-graphene-BN pressure sensing element to the ambient environment for 7 days, the relative resistance change in the pressure sensing element is only 3.1%, while that of the reference open-faced graphene device without the BN protection layers is 15.7%. Thus, this strategy is promising for fabricating practical graphene pressure sensors with improved performance and stability.

Mechanical, tribological and corrosion properties of CrBN films deposited by combined direct current and radio frequency magnetron sputtering

International Nuclear Information System (INIS)

Jahodova, Vera; Ding, Xing-zhao; Seng, Debbie H.L.; Gulbinski, W.; Louda, P.

2013-01-01

Cr–B–N films were deposited on stainless steel substrates by a combined direct current and radio frequency (RF) reactive unbalanced magnetron sputtering process using two elemental Cr and one compound BN targets. Boron content in the as-deposited films was qualitatively analyzed by time-of-flight secondary ion mass spectroscopy. Films' microstructure, mechanical and tribological properties were characterized by X-ray diffraction, nanoindentation and pin-on-disk tribometer experiments. Corrosion behavior of the Cr–B–N films was evaluated by electrochemical potentiodynamic polarization method in a 3 wt.% NaCl solution. All the films were crystallized into a NaCl-type cubic structure. At lower RF power applied on the BN target (≤ 600 W), films are relatively randomly oriented, and films' crystallinity increased with increasing RF power. With increasing RF power further (≥ 800 W), films became (200) preferentially oriented, and films' crystallinity decreased gradually. With incorporation of a small amount of boron atoms into the CrN films, hardness, wear- and corrosion-resistance were all improved evidently. The best wear and corrosion resistance was obtained for the film deposited with 600 W RF power applied on the BN target. - Highlights: • CrBN films deposited by direct current and radio frequency magnetron sputtering. • CrBN exhibited higher hardness, wear- and corrosion-resistance than pure CrN. • The best wear- and corrosion-resistant film was deposited with 600 W RF power

Development of natural circulation small and medium sized boiling water reactor: HSBWR-600

International Nuclear Information System (INIS)

Miki, Minoru; Horiuchi, Tetsuo; Yoshimoto, Yuichiro; Sumida, Isao; Murase, Michio; Akita, Minoru; Niino, Tsuyoshi

1988-01-01

In nuclear power generation, the development of large reactors has been promoted as the main energy source in Japan. However, world economy entered low growth age, and the growth of electric power demand slowed down. Accordingly, attention has been paid to the medium and small reactors that can cope with whatever needs by serializing their types in addition to the nuclear power plants of medium output matching to electric power demand. In order to cope with these new needs, the economical efficiency of medium and small reactors must be as close as possible to that of large reactors, and as the countermeasures to the demerits due to small size, those must be made into the plants having simplified systems and the safety easily acceptable to public. Hitachi Ltd. plans to develop the natural circulation type medium and small BWRs of 600 NWe output class, HSBWR-600, on the basis of the nuclear power plant technology based on the rich results of design and operation of BWRs obtained so far, and to rank them as one of the BWR series. The target of their development design, the circumstance of their development, the core design and the thermo-hydraulic characteristics, the reactor pressure vessel and in-core structures, the safety design, system design, building layout and the evaluation are reported. (Kako, I.)

TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

International Nuclear Information System (INIS)

Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

1983-01-01

TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

Synthesis and processing of nanostructured BN and BN/Ti composites

Science.gov (United States)

Horvath, Robert Steven

Superhard materials, such as cubic-BN, are widely used in machine tools, grinding wheels, and abrasives. Low density combined with high hardness makes c-BN and its composites attractive candidate materials for personnel and vehicular armor. However, improvements in toughness, and ballistic-impact performance, are needed to meet anticipated performance requirements. To achieve such improvements, we have targeted for development nanostructured c-BN, and its composites with Ti. Current research utilizes an experimental high pressure/high temperature (HPHT) method to produce these materials on a laboratory scale. Results from this work should transfer well into the industrial arena, utilizing high-tonnage presses used in the production of synthetic diamond and c-BN. Progress has been made in: (1) HPHT synthesis of cBN powder using Mg as catalyst; (2) HPHT consolidation of cBN powder to produce nanostructured cBN; (3) reactive-HPHT consolidation of mixed cBN/Ti powder to produce nanostructured Ti- or TiB2/TiN-bonded cBN; and (4) reactive-HPHT consolidation of mixed hBN/Ti powder to produce nanostructured Ti-bonded TiB2/TiN or TiB2/TiN. Even so, much remains to be done to lay a firm scientific foundation to enable the reproducible fabrication of large-area panels for armor applications. To this end, Rutgers has formed a partnership with a major producer of hard and superhard materials. The ability to produce hard and superhard nanostructured composites by reacting cBN or hBN with Ti under high pressure also enables multi-layered structures to be developed. Such structures may be designed to satisfy impedance-mismatch requirements for high performance armor, and possibly provide a multi-hit capability. A demonstration has been made of reactive-HPHT processing of multi-layered composites, consisting of alternating layers of superhard Ti-bonded cBN and tough Ti. It is noteworthy that the pressure requirements for processing Ti-bonded cBN, Ti-bonded TiB2/TiN, and their

Design considerations for economically competitive sodium cooled fast reactors

International Nuclear Information System (INIS)

Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

2009-01-01

The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, 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. (author)

Design concept of KALIMER-600

International Nuclear Information System (INIS)

Hahn, Dohee; Kim, Yeong-Il; Kim, Seong-O; Lee, Jae-Han; Lee, Yong-Bum

2005-01-01

KALIMER-600 is a pool-type sodium-cooled reactor loaded with U-TRU-10%Zr metal fuels generating the net electricity output of 600 MWe. In order to enhance the proliferation resistance, no blanket assemblies are loaded in the core. To suppress the high power peaking factor, some of the fuel rods are replaced with B 4 C rods and dummy rods. The heat transport system is comprised of two independent loops of IHTS and SGS and the safety-grade residual heat removal system, PDRC, is a completely passive system. Main features of the mechanical structure design of KALIMER-600 are the seismically isolated reactor building, the reduced total pipe length of the IHTS, the simplified reactor support, and the compact reactor internal structures. From the safety analyses, the KALIMER-600 design is verified to be capable of accommodating all the analyzed ATWS events. This self-regulation capability of the KALIMER-600 is mainly due to the inherent reactivity feedback mechanisms and completely passive PDRC system. (author)

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

International Nuclear Information System (INIS)

Orlov, V.V.

2001-01-01

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

Level-1 PSA to support the design of the KALIMER-600 Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Han, Sang Hoon; Kim, Tae-Woon; Jeong, Hae-Yong; Han, Seok Joong; Ahn, Kwang-Il; Yang, Joon-Eon

2012-01-01

A sodium-cooled fast reactor, KALIMER-600, is under development. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as a coolant. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing KALIMER-600 from the aspects of risk informed design. A preliminary level-1 internal full power PSA has been performed to evaluate the safety level and its applicability for the KALIMER-600 conceptual design. Various design alternatives are evaluated from the viewpoint of PSA in order to support the design of the KALIMER-600. Sensitivity studies are also performed to evaluate the assumptions made for the PSA. The applicability and weakness of the KALIMER-600 PSA are discussed. The technical issues to be solved in performing the PSA will be discussed. (authors)

DBE Analysis for KALIMER-600

International Nuclear Information System (INIS)

Ha, Kwi Seok; Jeong, Hae Young; Kwon, Young Min; Chang, Won Pyo; Lee, Yong Bum; Kim, Young II

2009-01-01

The SFR (Sodium Fast Reactor) which is being developed at KAERI (Korea Atomic Energy Research Institute) is currently divided into three types, such as, Advanced Concept 600 MWe break-even reactor and burner reactor and 1200 MWe break-even reactor. As a part of accidents analysis of the 600 MWe break-even reactor, 5 representative DBE's (Design Bases Events) are analyzed for the safety analysis. The 5 DBE's are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Pipe Break, and SBO (Station Black Out)

Role of Fugen HWR in Japan and design of a 600 MWe demonstration reactor

International Nuclear Information System (INIS)

Sawai, Sadamu.

1982-03-01

Fugen, a 165 MWe prototype of a heavy water-moderated, boiling light water-cooled reactor, has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work for a 600 MWe demonstration plant has been carried out since 1973. The important systems and components, such as pressure tube assemblies and control rod drive mechanism, are essentially the same as those of Fugen. However, some modification is made owing to the experience obtained in Fugen and LWrs. In the HWR Fugen, plutonium and uranium are effectively used, and plutonium makes the coolant void reactivity more negative, which results in the increase of the stability and safety of the reactor. On August 4, 1981, the ad hoc committee submitted the final report to the Japanese Atomic Energy Commission, in which the construction of a 600 MWe demonstration plant was recommended. As for the research and development on reactor safety, coolant leak detectors, the performance of ECCS, and safety design codes are enumerated. Since 1965, mixed oxide fuel has been developed, and 168 fuel assemblies were loaded in Fugen, but failure did not occur. (Kako, I.)

Spin transport in two-layer-CVD-hBN/graphene/hBN heterostructures

Science.gov (United States)

Gurram, M.; Omar, S.; Zihlmann, S.; Makk, P.; Li, Q. C.; Zhang, Y. F.; Schönenberger, C.; van Wees, B. J.

2018-01-01

We study room-temperature spin transport in graphene devices encapsulated between a layer-by-layer-stacked two-layer-thick chemical vapor deposition (CVD) grown hexagonal boron nitride (hBN) tunnel barrier, and a few-layer-thick exfoliated-hBN substrate. We find mobilities and spin-relaxation times comparable to that of SiO2 substrate-based graphene devices, and we obtain a similar order of magnitude of spin relaxation rates for both the Elliott-Yafet and D'Yakonov-Perel' mechanisms. The behavior of ferromagnet/two-layer-CVD-hBN/graphene/hBN contacts ranges from transparent to tunneling due to inhomogeneities in the CVD-hBN barriers. Surprisingly, we find both positive and negative spin polarizations for high-resistance two-layer-CVD-hBN barrier contacts with respect to the low-resistance contacts. Furthermore, we find that the differential spin-injection polarization of the high-resistance contacts can be modulated by dc bias from -0.3 to +0.3 V with no change in its sign, while its magnitude increases at higher negative bias. These features point to the distinctive spin-injection nature of the two-layer-CVD-hBN compared to the bilayer-exfoliated-hBN tunnel barriers.

Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

International Nuclear Information System (INIS)

Choi, Seok Ki; Kim, Seong O

2011-01-01

A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

Settlement substantiation of the passive devices shutdown fast reactors by trip the absorbing rod in case of anticipated accident

International Nuclear Information System (INIS)

Portianoy, A.G.; Serdun, E.N.; Sorokin, A.P.; Uhov, V.A.; Egorov, V.S.

2000-01-01

Results of improvement of the passive device shutdown fast reactors BN-600 (PDSR) are considered. The device works (lets off a neutron absorber) at increase of coolant temperature above 660 deg. C (650 deg. C). The PDSR working element represents a design of a sylphon-container type, filled with aluminium (magnesium) and operates (extended) under melting it at the expense of energy of a compressed high-temperature spring, and/or increases of a volume (6% of aluminium) at melting, and/or increases of a volume at further growth of a temperature. Account of the characteristics of PDSR working elements is carried out. Mathematical models, describing dependence of the basic of the characteristics (sluggishness, size of lengthening) from the constructive factors and modes of anticipated accident, are received. Is shown, that the PDSR characteristics provide an emergency stop of the reactor BN-600 in a case of a heaviest anticipated accident prior to the beginning sodium boiling in a core. The developed PDSR have a number of advantages before known, for example, magnetic with a Curie point, first of all, at the expense of significant efforts generation, multichannels of operation and weak dependence on the operational factors, first of all, neutron fluence. (author)

Study of short-time mechanical properties changes for BN-350 reactor spent fuel assemblies jacket material from vacancy swelling

International Nuclear Information System (INIS)

Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.L.; Golovin, S.V.; Lambert, D.

1999-01-01

Variations of mechanical properties (ultimate strength and limit of plasticity) for irradiated stainless steels, materials of BN-350 reactor cased fuel assemblies tubes, namely: 12X18H10T MTO, 08X16H11M3 MTO, 10X17H13M2T, 12X13M2BRF from vacancy swelling and neutron damaging doze have been studied. Flat samples cut out from hexagonal fuel assemblies casing were tested. The data on casing profilometry, and also the results from hydrostatic weighing of steel samples, were used to evaluate swelling. All measurements and testing were made at temperature 25 degrees C

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-11-01

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.

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

International Nuclear Information System (INIS)

1977-11-01

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

Anisotropic Effective Mass, Optical Property, and Enhanced Band Gap in BN/Phosphorene/BN Heterostructures.

Science.gov (United States)

Hu, Tao; Hong, Jisang

2015-10-28

Phosphorene is receiving great research interests because of its peculiar physical properties. Nonetheless, the phosphorus has a trouble of degradation due to oxidation. Hereby, we propose that the electrical and optical anisotropic properties can be preserved by encapsulating into hexagonal boron nitride (h-BN). We found that the h-BN contributed to enhancing the band gap of the phosphorene layer. Comparing the band gap of the pristine phosphorene layer, the band gap of the phosphorene/BN(1ML) system was enhanced by 0.15 eV. It was further enhanced by 0.31 eV in the BN(1ML)/phosphorene/BN(1ML) trilayer structure. However, the band gap was not further enhanced when we increased the thickness of the h-BN layers even up to 4 MLs. Interestingly, the anisotropic effective mass and optical property were still preserved in BN/phosphorene/BN heterostructures. Overall, we predict that the capping of phosphorene by the h-BN layers can be an excellent solution to protect the intrinsic properties of the phosphorene.

Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

International Nuclear Information System (INIS)

Huang Xueqing; Zhang Senru

1999-01-01

Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

BN interphase in composite materials with nicalon Si-C-O fibers and with vitro ceramic matrix of MAS type; L`interphase BN dans les materiaux composites a fibres Si-C-O nicalon et a matrice vitroceramique de type MAS

Energy Technology Data Exchange (ETDEWEB)

Ricca, N

1994-03-14

BN has been suggested as an interphase in silica-based glass-ceramic matrix composites with a view to use these materials in oxidizing atmospheres at medium or high temperatures. The matrix had a boron-doped MAS (MgO-Al{sub 2}O{sub 3}-SiO{sub 2}) composition and was prepared from an hydrosol precursor. Pseudo-ID composites were prepared according to a sol impregnations/calcination/hot-pressing route. Chemical and microstructural characterizations of the fiber/matrix interfacial area were conducted by mean of TEM/EELS and AES analyses. The efficiency of BN as a coupling interphase for this particular composite system was successfully demonstrated through tensile tests performed on either as-processed or aged specimens (100 hours at 1000 deg C in air or under argon). In addition, composites maintained in air at 600 deg C, 800 deg C and 900 deg C while simultaneously loaded did not fail after 150 hours or more. Thus, a BN interphase appeared to be compatible with an oxidizing environment (i.e. the oxide matrix and/or air from 600 to 1000 deg C) and should therefore successfully replace the usual carbon interphase at least for use at medium temperatures. (author)

Bias induced modulation of electrical and thermal conductivity and heat capacity of BN and BN/graphene bilayers

Energy Technology Data Exchange (ETDEWEB)

Chegel, Raad, E-mail: Raad.chegel@gmail.com

2017-04-15

By using the tight binding approximation and Green function method, the electronic structure, density of state, electrical conductivity, heat capacity of BN and BN/graphene bilayers are investigated. The AA-, AB{sub 1}- and AB{sub 2}- BN/graphene bilayers have small gap unlike to BN bilayers which are wide band gap semiconductors. Unlike to BN bilayer, the energy gap of graphene/BN bilayers increases with external field. The magnitude of the change in the band gap of BN bilayers is much higher than the graphene/BN bilayers. Near absolute zero, the σ(T) is zero for BN bilayers and it increases with temperature until reaches maximum value then decreases. The BN/graphene bilayers have larger electrical conductivity larger than BN bilayers. For both bilayers, the specific heat capacity has a Schottky anomaly.

Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

International Nuclear Information System (INIS)

Moraleda Lopez, A.

1997-01-01

The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

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

International Nuclear Information System (INIS)

1978-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1978-09-01

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.

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

International Nuclear Information System (INIS)

Moiseyev, A.V.

2008-01-01

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 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)

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)

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)

2004-01-01

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

Role of Fugen-HWR in Japan and design of a 600 MWe demonstration reactor

International Nuclear Information System (INIS)

Sawai, S.

1982-01-01

Fugen, a 165 MWe prototype of a heavy water moderated boiling light water cooled reactor; has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work of the 600 MWe demonstration plant has been carried out since 1973. Important system and components, such as pressure tube assemblies, control rod drive mechanism, etc., are essentially the same as those of Fugen. Some modifications, however, are made especially from the stand point of experiences In the Fugen-HWR, plutonium and uranium would be effectively used; and plutonium could make the coolant void reactivity more negative which would give good results in increasing the reactor stability and safety. On the other hand, nuclear power plants are mainly consisted of LWRs in Japan. Considering the above situations, the Fugen-HWR, coupled with LWRs, is now considered in Japan to contribute to our energy security by using plutonium and depleted uranium extracted from spent fuels of LWRs: thereby reducing the demands On August 4, 1981, the ad hoc committee on the 600 MWe demonstration Fugen-HWR submitted the final report to the Japan AEC, after having had discussions and evaluations. In the report, the ad hoc committee recommended to build the 600 MWE demonstration plant with appropriate supports of the Government. The Japan AEC will be expected to make her decision on the program in the near future. As for the reactor safety R and C, development has been stressed on coolant leak detectors and ECCS performances or Since 1965, many development works have been done for mixed oxide fuel assemblies, both for establishment of the fabrication technology and for clarification of irradiation performances. 196 mixed oxide fuel assemblies have been manufactured for Fugen. 168 of them were loaded and 92 were withdrawn. No fuel has been failured yet. (author)

Densification Behavior of BN-added UO2

International Nuclear Information System (INIS)

Rhee, Young Woo; Kim, Keonsik; Kim, Dong Joo; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho

2013-01-01

Local wall thinning in pipelines affects the structural integrity of industries like nuclear power plants (NPPs). In the present study a pulsed eddy current (PEC) technology to detect the wall thing of carbon steel pipe covered with insulation is developed. Boron is commercially used as a neutron absorber fuel. A neutron absorber fuel is burned out or depleted during reactor operation. Westinghouse have been produced the Integral Fuel Burnable Absorber (IFBA) which is enriched UO 2 fuel pellets with a thin coating of zirconium diboride (ZrB 2 ) on the outer surface. Standard sintered fuel pellets are sputter coated with ZrB 2 . It is known that IFBA fuel can incur 20% to 30% additional fabrication costs. Boron-dispersed UO 2 fuel pellet made by the conventional pressing and sintering process of a powder mixture of UO 2 and B compound might be more cost-effective than IFBAs. M. G. Andrew et al. tried to sinter boron-dispersed UO 2 green pellet. However, they reported that boron-dispersed UO 2 fuel pellet is very difficult to be fabricated with a sufficient level of boron retention and high sintered density (greater than 90 % of theoretical density) because of the volatilization of boron oxide. We have investigated the densification behavior of mixtures of UO 2 and various boron compounds, such as B 4 C, BN, TiB 2 , ZrB 2 , SiB 6 , and HfB 2 . Boron compounds seemed to act as a sintering additive for UO 2 at a certain low temperature range. In this study, the densification behavior of BN-added UO 2 pellet has been investigated by sintering green pellets of a mixture of UO 2 powder and BN powder in H 2 atmosphere. A high density BN-added UO 2 pellet can be fabricated after sintering at 1200 .deg. C for more than 1 h in a H 2 atmosphere. The sintered density of BN-added UO 2 pellet can be increased up to about 95 %TD

Analysis and application of a simulator of a nuclear reactor AP-600; Analisis y aplicacion de un simulador de un reactor nuclear AP-600

Energy Technology Data Exchange (ETDEWEB)

Medina S, V. S. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Salazar S, E., E-mail: medina_victor@comunidad.unam.mx [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (MX)

2011-11-15

In front of the resurgence of interest in the nuclear power production, several national organizations have considered convenient to have highly specialized human resources in the technologies of nuclear reactors of III + and IV generation. For this task, the intensive and extensive applications of the computation should been considered, as the virtual instrumentation. The present work analyzes the possible applications of a nuclear simulator provided by the IAEA with base in the design of the reactor AP-600, using a focusing of modular model developed in FORTRAN. One part of the work that was made with the simulator includes the evaluation of 21 transitory events of operation, including the recreation of the accident happened in the nuclear power plant of Three Mile Island in 1979, comparing the actions flow and the answer of the systems under the intrinsic security of a III + generation reactor. The impact that had the mentioned accident was analyzed in the growing of the nuclear energy sector and in the public image with regard to the nuclear power plants. An application for this simulator was proposed, its use as tool for the instruction in the nuclear engineering courses using it to observe the operation of the different security systems and its interrelation inside the power plant as well as a theoretical/practical approach for the student. (Author)

(Fuzzy) Ideals of BN-Algebras

Science.gov (United States)

Walendziak, Andrzej

2015-01-01

The notions of an ideal and a fuzzy ideal in BN-algebras are introduced. The properties and characterizations of them are investigated. The concepts of normal ideals and normal congruences of a BN-algebra are also studied, the properties of them are displayed, and a one-to-one correspondence between them is presented. Conditions for a fuzzy set to be a fuzzy ideal are given. The relationships between ideals and fuzzy ideals of a BN-algebra are established. The homomorphic properties of fuzzy ideals of a BN-algebra are provided. Finally, characterizations of Noetherian BN-algebras and Artinian BN-algebras via fuzzy ideals are obtained. PMID:26125050

Integrated thermal analysis of top-shield and reactor vault of Indian FBR-600

International Nuclear Information System (INIS)

Rajendrakumar, M.; Velusamy, K.; Selvaraj, P.

2015-01-01

The design for next generation fast breeder reactors (FBR-600) has been commenced with enhanced safety and improved economy as the main targets. The Top Shield (TS) of Prototype Fast Breeder Reactor (PFBR) is a box type structure consisting of Roof Slab (RS), Small Rotatable Plug (SRP), and Large Rotatable Plug (LRP). The large box type structure with many penetrations posed difficulties during manufacturing. Because of the required high load carrying capabilities, a dome shaped thick plate roof slab is conceived for FBR-600. Main Vessel (MV) which holds the primary sodium and associated components is welded to the RS through a triple joint. Reactor vault (RV) is a thick concrete structure which supports MV and Safety Vessel (SV). The temperature of RV concrete has to be less than 338 K (65°C) under normal operating heat loads (full and part load conditions) and less than 363 K (90°C) under Safety Grade Decay Heat Removal (SGDHR) conditions with one cooling loop in service. The temperature in the component penetrations of the RS should be greater than 120°C to avoid sodium aerosol deposition. Similarly, the temperature of the LRP and SRP has to be ∼120°C to protect the elastomeric seals provided to these structures. Further, the heat load to RV transferred by direct conduction by roof slab support has to be minimum. To meet these conflicting thermal requirements, detailed multi-physics CFD calculations have been performed to finalize, (i) the insulation requirements on the top of roof slab, (ii) number and position of reflective insulation plates below the bottom plate of roof slab/rotating plugs, (iii) air flow rate for various zones of the top shield and (iv) water flow rate and pitch of water cooling pipes for the reactor vault. (author)

Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

International Nuclear Information System (INIS)

Guidez, J.; Saturnin, A.

2016-01-01

In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

Effect of magnesium aluminum silicate glass on the thermal shock resistance of BN matrix composite ceramics

NARCIS (Netherlands)

Cai, Delong; Jia, Dechang; Yang, Zhihua; Zhu, Qishuai; Ocelik, Vaclav; Vainchtein, Ilia D.; De Hosson, Jeff Th M.; Zhou, Yu

The effects of magnesium aluminum silicate (MAS) glass on the thermal shock resistance and the oxidation behavior of h-BN matrix composites were systematically investigated at temperature differences from 600 degrees C up to 1400 degrees C. The retained strength rate of the composites rose with the

AP600 design certification thermal hydraulics testing and analysis

Energy Technology Data Exchange (ETDEWEB)

Hochreiter, L.E.; Piplica, E.J.

1995-09-01

Westinghouse Electric Corporation, in conjunction with the Department of Energy and the Electric Power Research Institute, have been developing an advanced light water reactor design; the AP600. The AP600 is a 1940 Mwt, 600Mwe unit which is similar to a Westinghouse two-loop Pressurized Water Reactor. The accumulated knowledge on reactor design to reduce the capital costs, construction time, and the operational and maintenance cost of the unit once it begins to generate electrical power. The AP600 design goal is to maintain an overall cost advantage over fossil generated electrical power.

Research Update: Direct conversion of h-BN into pure c-BN at ambient temperatures and pressures in air

Directory of Open Access Journals (Sweden)

Jagdish Narayan

2016-02-01

Full Text Available We report a direct conversion of hexagonal boron nitride (h-BN into pure cubic boron nitride (c-BN by nanosecond laser melting at ambient temperatures and atmospheric pressure in air. According to the phase diagram, the transformation from h-BN into c-BN can occur only at high temperatures and pressures, as the hBN-cBN-Liquid triple point is at 3500 K/9.5 GPa. Using nanosecond laser melting, we have created super undercooled state and shifted this triple point to as low as 2800 K and atmospheric pressure. The rapid quenching from super undercooled state leads to formation of super undercooled BN (Q-BN. The c-BN phase is nucleated from Q-BN depending upon the time allowed for nucleation and growth.

Unusual occurrences during the whole operation of BN-250 NPP

International Nuclear Information System (INIS)

Andropenkov, S.

2000-01-01

Unusual occurrences during the whole operation BN-350 NPP. 1. Oil ingress in high pressure receiver for the not reveled reason, 12.05.1994. 2. lncrease of water radioactivity of circulating water supply system due to heat exchanger leak of spent fuel assembly washing out system, 17.09.1993. 3. Lack of passableness of sodium drain header of primary circuit reveled during inspection on scheduled preventative maintenance, 28.11.1996. 4. Destruction of the blow-off line of MCP-6 due to corrosion damage of the pipeline while unit was being operated at rated power, 23.04.1993. 5. Lack of passableness of blow-down pipeline connecting reactor gas cover with gas-type pressurizer while unit was being operated at rated power, 17.11.1994. 6. Sodium ingress in blow-down pipeline of loop-5 intermediate heat exchanger while loop-5 was being fed of sodium during scheduled preventative maintenance, 27.06.1994. 7. Resistance deterioration of electro heating zones of loop-4 due to heat exchanger leak and water ingress in air-pipeline of primary circuit boxes recirculating air system, 02.05.1997. 8. Resistance deterioration of electro heating zones of sodium drain header of secondary circuit was sopped in the water for the extinguishing the fire of blowing ventilation oil-strainer, 23.12.1994. 9. Sodium ingress in gas-type pressurizer through pipeline of primary sodium cleanup system and blow-down pipeline of failed MCP-2 while primary sodium cleanup system was being connected to the primary circuit, 17.08.1976. As a rule, the main reactor systems are scrutinized more carefully than the auxiliary reactor systems and the order actions are existed for eliminating and mitigating of consequences of main reactor system fails. Therefore the auxiliary reactor system fails may impact on the main reactor systems through places of its contact in significant measure. The influence of auxiliary reactor system fails on main reactor systems and its possible consequences for behavior of the main

Tribological properties of epoxy composite coatings reinforced with functionalized C-BN and H-BN nanofillers

Science.gov (United States)

Yu, Jingjing; Zhao, Wenjie; Wu, Yinghao; Wang, Deliang; Feng, Ruotao

2018-03-01

A series of epoxy resin (EP) composite coatings reinforced with functionalized cubic boron nitride (FC-BN) and functionalized hexagonal boron nitride (FH-BN) were fabricated successfully on 316L stainless steel by hand lay-up technique. The structure properties were characterized by Fourier transform infrared spectroscopy (FTIR), X-ray photoelectron spectroscopy (XPS) and X-ray diffraction (XRD). The morphologies were characterized by atomic force microscopy (AFM), scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Moreover, UMT-3 tribometer and surface profiler were used to investigate tribological behaviors of as-prepared composite coatings under dry friction and seawater conditions respectively. The results demonstrated that the presence of FC-BN or FH-BN fillers could greatly decrease the friction coefficient (COF) and wear rate of epoxy, in addition, composite coatings possess better tribological properties under seawater condition which was attributed to the lubricating effect of seawater. Moreover, FC-BN endows the composite coatings the highest wear resistance, and FH-BN /EP composite coatings exhibited the best friction reduction performance which is attributed to the self-lubricating performance of lamella structure for FH-BN sheet.

Structure and creep of Russian reactor steels with a BCC structure

Science.gov (United States)

Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

2017-05-01

The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

Concept of the plant for the BN-800 fast reactor fuel recycling with application of pyro-process and vibro-packing technology

International Nuclear Information System (INIS)

Bychkov, A.V.; Skiba, O.V.; Mayorshin, A.A.; Demidova, L.S.; Kormilitzyna, L.A.; Ishunin, V.S.

2000-01-01

The conception of Plant was developed for MOX-fuel recycle at two BN-800 type fast reactors by pyrochemical reprocessing of irradiated nuclear fuel (INF) and production of vibro-pac fuel pins and SA. INF production process and stages of pyrochemical reprocessing were analyzed. Starting materials were chosen. Characteristics of irradiated SA and requirements for finished products were defined. Volumes of production were estimated. Procedure of waste management was defined. The following description was made: (1) general flow sheet of fuel recycling and partial schemes of single reprocessing; (2) composition of production process equipment; (3) arrangement of production process equipment; (4) lay out of Plant building and engineering communications. Principle economical assessments were made for production under design. (authors)

Acoustic control system BN-350. Explanatory note

International Nuclear Information System (INIS)

1982-02-01

A description of the acoustic system developed to control boiling in the active zone of the BN 350 reactor is given together with the corresponding technical characteristics. The results of experiments and calculations which confirm the validity of the solutions adopted are discussed. Theoretical calculations on the boiling process in the duct are reported together with details on the fast diagnostic system. A means for localizing the onset of boiling is also given, possible error being taken into consideration. The special features of the passive acoustic diagnostic method used to study boiling are described and schemas of the anciliary equipment presented [fr

Local flow distribution analysis inside the reactor pools of KALIMER-600 and PDRC performance test facility

International Nuclear Information System (INIS)

Jeong, Ji Hwan; Hwang, Seong Won; Choi, Kyeong Sik

2010-05-01

In the study, 3-dimensional thermal hydraulic analysis was carried out focusing on the thermal hydraulic behavior inside the reactor pools for both KALIMER-600 and one-fifth scale-down test facility. STAR-CD, one of the commercial CFD codes, was used to analyze 3-dimensional incompressible steady-state thermal hydraulic behavior in both designs of KALIMER-600 and the scale-down test facility. In the KALIMER-600 CFD analysis, the pressure drops in the core and IHX gave a good agreement within 1% error range. It was found that the porous media model was appropriate to analyze the pressure distribution inside reactor core and IHX. Also, a validation analysis showed the pressure drop through the porous media under the condition of 80% flow rate and thermal power was calculated 64% less than in 100% condition giving a physically reasonable analytic result. Since the temperatures in the hot-side pool and cold-side pool were estimated to be very close to 540 and 390 .deg. C specified on the design values respectively, the CFD models of heat source and sink was confirmed. Through the study, the methodology of 3-dimensional CFD analysis about KALIMER-600 has been established and proven. Performed with the methodology, the analysis data such as flow velocity, temperature and pressure distribution were compared by normalizing those data for the actual sized modeling and scale-down modeling. As a result, the characteristics of thermal hydraulic behavior were almost identical for the actual sized modeling and scale-down modeling and the similarity scaling law used in the design of the sodium test facility by KAERI was found to be correct

International collaborations about fuel studies for reactor recycling of military quality plutonium

International Nuclear Information System (INIS)

Bernard, H.; Chaudat, J.P.

1997-01-01

In November 1992, an agreement was signed between the French and Russian governments to use in Russia and for pacific purposes the plutonium recovered from the Russian nuclear weapons dismantling. This plutonium will be transformed into mixed oxide fuels (MOX) for nuclear power production. The French Direction of Military Applications (DAM) of the CEA is the operator of the French-Russian AIDA program. The CEA Direction of Fuel Cycle (DCC) and Direction of Nuclear Reactors (DRN) are involved in the transformation of metallic plutonium into sinterable oxide powder for MOX fuel manufacturing. The Russian TOMOX (Treatment of MOX powder Metallic Objects) and DEMOX (MOX Demonstration) plants will produce the MOX fuel assemblies for the 4 VVER 1000 reactors of Balakovo and the fast BN 600 reactor. The second part of the program will involve the German Siemens and GRS companies for the safety studies of the reactors and fuel cycle plants. The paper gives also a brief analysis of the US policy concerning the military plutonium recycling. (J.S.)

The effect of prior deformation on stress corrosion cracking growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water

International Nuclear Information System (INIS)

Yamazaki, Seiya; Lu Zhanpeng; Ito, Yuzuru; Takeda, Yoichi; Shoji, Tetsuo

2008-01-01

The effect of prior deformation on stress corrosion cracking (SCC) growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water environment is studied. The prior deformation was introduced by welding procedure or by cold working. Values of Vickers hardness in the Alloy 600 weld heat-affected zone (HAZ) and in the cold worked (CW) Alloy 600 materials are higher than that in the base metal. The significantly hardened area in the HAZ is within a distance of about 2-3 mm away from the fusion line. Electron backscatter diffraction (EPSD) results show significant amounts of plastic strain in the Alloy 600 HAZ and in the cold worked Alloy 600 materials. Stress corrosion cracking growth rate tests were performed in a simulated pressurized water reactor primary water environment. Extensive intergranular stress corrosion cracking (IGSCC) was found in the Alloy 600 HAZ, 8% and 20% CW Alloy 600 specimens. The crack growth rate in the Alloy 600 HAZ is close to that in the 8% CW base metal, which is significantly lower than that in the 20% CW base metal, but much higher than that in the as-received base metal. Mixed intergranular and transgranular SCC was found in the 40% CW Alloy 600 specimen. The crack growth rate in the 40% CW Alloy 600 was lower than that in the 20% CW Alloy 600. The effect of hardening on crack growth rate can be related to the crack tip mechanics, the sub-microstructure (or subdivision of grain) after cross-rolling, and their interactions with the oxidation kinetics

Preliminary analysis of the proposed BN-600 benchmark core

International Nuclear Information System (INIS)

John, T.M.

2000-01-01

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)

The use of waveguide acoustic probes for void fraction measurement in the evaporator of BN-350-Type reactor

Energy Technology Data Exchange (ETDEWEB)

Melnikov, V.I.; Nigmatulin, B.I.

1995-09-01

The present paper deals with some results of the experimental studies which have been carried out to investigate the steam generation dynamics in the Field tubes of sodium-water evaporators used in the BN-350 reactors. The void fraction measurements have been taken with the aid of waveguide acoustic transducers manufactured in accordance with a specially designed technology (waveguide acoustic transducers-WAT technology). Presented in this paper also the transducer design and calibration methods, as well as the diagram showing transducers arrengment in the evaporator. The transducers under test featured a waveguide of about 4 m in length and a 200-mm long sensitive element (probe). Besides, this paper specifies the void fraction data obtained through measurements in diverse points of the evaporator. The studies revealed that the period of observed fluctuations in the void fraction amounted to few seconds and was largely dependent on the level of water in the evaporator.

Vibrational Properties of h-BN and h-BN-Graphene Heterostructures Probed by Inelastic Electron Tunneling Spectroscopy.

Science.gov (United States)

Jung, Suyong; Park, Minkyu; Park, Jaesung; Jeong, Tae-Young; Kim, Ho-Jong; Watanabe, Kenji; Taniguchi, Takashi; Ha, Dong Han; Hwang, Chanyong; Kim, Yong-Sung

2015-11-13

Inelastic electron tunneling spectroscopy is a powerful technique for investigating lattice dynamics of nanoscale systems including graphene and small molecules, but establishing a stable tunnel junction is considered as a major hurdle in expanding the scope of tunneling experiments. Hexagonal boron nitride is a pivotal component in two-dimensional Van der Waals heterostructures as a high-quality insulating material due to its large energy gap and chemical-mechanical stability. Here we present planar graphene/h-BN-heterostructure tunneling devices utilizing thin h-BN as a tunneling insulator. With much improved h-BN-tunneling-junction stability, we are able to probe all possible phonon modes of h-BN and graphite/graphene at Γ and K high symmetry points by inelastic tunneling spectroscopy. Additionally, we observe that low-frequency out-of-plane vibrations of h-BN and graphene lattices are significantly modified at heterostructure interfaces. Equipped with an external back gate, we can also detect high-order coupling phenomena between phonons and plasmons, demonstrating that h-BN-based tunneling device is a wonderful playground for investigating electron-phonon couplings in low-dimensional systems.

Development of devices for handling with BN-350 radioactive waste

International Nuclear Information System (INIS)

Iksanov, A.G.; Pustobaev, S.N.; Shirobokov, Yu.P.; Pugachyev, G.P.; Baldov, A.N.; Tikhomirov, L.N.; Tkachenko, V.V.; Tazhibayeva, I.L.; Klepikov, A.Kh.; Romanenko, O.G.; Kenzhin, E.A.; Yakovlev, V.V.; Khametov, S.; Kalinkin, V.L.; Skvortsov, A.I.; Dmitriev, S.A.; Arustamov, A.E.; Zelenski, D.I.; Serebrennikov, Yu.A.

2010-01-01

The package of activity performed proves the correctness of the concept accepted by the Government of the Republic of Kazakhstan on the BN-350 decommissioning (three successive steps above) targeted at minimization of cost, exposure and amount of radioactive waste. Decommissioning of the high power fast breeder reactor plant is carried out for the first time and therefore the normative documents and design decisions elaborated, accepted technologies and estimation of capital expenditure and maintenance costs may enrich the database and serve as orientation for decommissioning of similar units. According to the concept accepted the BN-350 decommissioning is the process of top level of complexity that is characterized with the requirement of concurrent execution of a large scope of work by means of international teams from Kazakhstan, Russia, USA, EC, etc. Such approach needs the creation of modern effective organization schemes of interfaces and management of the Projects and will be further used in other complicated Projects

AP600 - an ALWR conceptual design

International Nuclear Information System (INIS)

Bruce, R.A.; Vijuk, R.P.

1988-01-01

The Electric Power Research Institute is spearheading an effort to develop utility requirements for the Advanced Light Water Reactor (ALWR) plants which will become the next generation nuclear power plants for the U.S. This EPRI ALWR Program involves utilities, the U.S. Department of Energy, the U.S. Nuclear Regulatory Commission, and various industry suppliers. The ALWR Program is aimed at ALWR plants which incorporate step improvements in safety, reliability, operability and power generation costs. As part of the ALWR efforts, a Westinghouse team is conducting conceptual design development of a PWR plant design called the AP600, reflecting advanced passive safety features and the chosen 600 MWe plant output. The AP600 conceptual design provides significant improvements while employing proven component technology. This paper describes the basic reactor and primary coolant system features, the passive safety system features, and plant arrangement/construction features of AP600

Cellulose nanofibrils (CNF) filled boron nitride (BN) nanocomposites

Energy Technology Data Exchange (ETDEWEB)

Sulaiman, Hanisah Syed; Hua, Chia Chin; Zakaria, Sarani [School of Applied Physic, Faculty of Science and Technology, Universiti Kebangsaan Malaysia.43600 Bangi, Selangor (Malaysia)

2015-09-25

In this study, nanocomposite using cellulose nanofibrils filled with different percentage of boron nitride (CNF-BN) were prepared. The objective of this research is to study the effect of different percentage of BN to the thermal conductivity of the nanocomposite produced. The CNF-BN nanocomposite were characterization by FT-IR, SEM and thermal conductivity. The FT-IR analysis of the CNF-BN nanocomposite shows all the characteristic peaks of cellulose and BN present in all samples. The dispersion of BN in CNF were seen through SEM analysis. The effect of different loading percentage of BN to the thermal conductivity of the nanocomposite were also investigated.

Methodology for Applying Cyber Security Risk Evaluation from BN Model to PSA Model

Energy Technology Data Exchange (ETDEWEB)

Shin, Jin Soo; Heo, Gyun Young [Kyung Hee University, Youngin (Korea, Republic of); Kang, Hyun Gook [KAIST, Dajeon (Korea, Republic of); Son, Han Seong [Joongbu University, Chubu (Korea, Republic of)

2014-08-15

There are several advantages to use digital equipment such as cost, convenience, and availability. It is inevitable to use the digital I and C equipment replaced analog. Nuclear facilities have already started applying the digital system to I and C system. However, the nuclear facilities also have to change I and C system even though it is difficult to use digital equipment due to high level of safety, irradiation embrittlement, and cyber security. A cyber security which is one of important concerns to use digital equipment can affect the whole integrity of nuclear facilities. For instance, cyber-attack occurred to nuclear facilities such as the SQL slammer worm, stuxnet, DUQU, and flame. The regulatory authorities have published many regulatory requirement documents such as U.S. NRC Regulatory Guide 5.71, 1.152, IAEA guide NSS-17, IEEE Standard, and KINS Regulatory Guide. One of the important problem of cyber security research for nuclear facilities is difficulty to obtain the data through the penetration experiments. Therefore, we make cyber security risk evaluation model with Bayesian network (BN) for nuclear reactor protection system (RPS), which is one of the safety-critical systems to trip the reactor when the accident is happened to the facilities. BN can be used for overcoming these problems. We propose a method to apply BN cyber security model to probabilistic safety assessment (PSA) model, which had been used for safety assessment of system, structure and components of facility. The proposed method will be able to provide the insight of safety as well as cyber risk to the facility.

Methodology for Applying Cyber Security Risk Evaluation from BN Model to PSA Model

International Nuclear Information System (INIS)

Shin, Jin Soo; Heo, Gyun Young; Kang, Hyun Gook; Son, Han Seong

2014-01-01

There are several advantages to use digital equipment such as cost, convenience, and availability. It is inevitable to use the digital I and C equipment replaced analog. Nuclear facilities have already started applying the digital system to I and C system. However, the nuclear facilities also have to change I and C system even though it is difficult to use digital equipment due to high level of safety, irradiation embrittlement, and cyber security. A cyber security which is one of important concerns to use digital equipment can affect the whole integrity of nuclear facilities. For instance, cyber-attack occurred to nuclear facilities such as the SQL slammer worm, stuxnet, DUQU, and flame. The regulatory authorities have published many regulatory requirement documents such as U.S. NRC Regulatory Guide 5.71, 1.152, IAEA guide NSS-17, IEEE Standard, and KINS Regulatory Guide. One of the important problem of cyber security research for nuclear facilities is difficulty to obtain the data through the penetration experiments. Therefore, we make cyber security risk evaluation model with Bayesian network (BN) for nuclear reactor protection system (RPS), which is one of the safety-critical systems to trip the reactor when the accident is happened to the facilities. BN can be used for overcoming these problems. We propose a method to apply BN cyber security model to probabilistic safety assessment (PSA) model, which had been used for safety assessment of system, structure and components of facility. The proposed method will be able to provide the insight of safety as well as cyber risk to the facility

Analysis and application of a simulator of a nuclear reactor AP-600

International Nuclear Information System (INIS)

Medina S, V. S.; Salazar S, E.

2011-11-01

In front of the resurgence of interest in the nuclear power production, several national organizations have considered convenient to have highly specialized human resources in the technologies of nuclear reactors of III + and IV generation. For this task, the intensive and extensive applications of the computation should been considered, as the virtual instrumentation. The present work analyzes the possible applications of a nuclear simulator provided by the IAEA with base in the design of the reactor AP-600, using a focusing of modular model developed in FORTRAN. One part of the work that was made with the simulator includes the evaluation of 21 transitory events of operation, including the recreation of the accident happened in the nuclear power plant of Three Mile Island in 1979, comparing the actions flow and the answer of the systems under the intrinsic security of a III + generation reactor. The impact that had the mentioned accident was analyzed in the growing of the nuclear energy sector and in the public image with regard to the nuclear power plants. An application for this simulator was proposed, its use as tool for the instruction in the nuclear engineering courses using it to observe the operation of the different security systems and its interrelation inside the power plant as well as a theoretical/practical approach for the student. (Author)

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

International Nuclear Information System (INIS)

Friess, Friederike Renate

2017-01-01

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

Energy Technology Data Exchange (ETDEWEB)

Friess, Friederike Renate

2017-07-12

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Model Based Cyber Security Analysis for Research Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)

2013-07-01

The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.

Model Based Cyber Security Analysis for Research Reactor Protection System

International Nuclear Information System (INIS)

Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung; Son, Hanseong

2013-01-01

The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN

Westinghouse AP600 advanced nuclear plant design

International Nuclear Information System (INIS)

Gangloff, W.

1999-01-01

As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

Study on thermal conductive BN/novolac resin composites

International Nuclear Information System (INIS)

Li, Shasha; Qi, Shuhua; Liu, Nailiang; Cao, Peng

2011-01-01

Highlights: → Boron nitride (BN) particles were used to modify novolac resin. → BN particles were pretreated by γ-aminopropyltriethoxysilane. → The thermal conductivity trend of composite almost agrees with the predicted data from the Maxwell-Eucken model. → At BN concentration of 80 wt.%, thermal conductivity value of composite is 4.5 times that of pure novolac resin. → Combined use of the larger and smaller particles with a mass ratio of 1:2 provides the composites with the maximum thermal conductivity among the testing systems. → The composite thermal property also increases with an increase in the BN concentration. - Abstract: In this study, γ-aminopropyltriethoxysilane-treated boron nitride (BN) particles were used to modify novolac resin. The effect of varying the BN concentration, particle size, and hybrid BN fillers with the binary particle size distribution on the thermal conductivity of the composites was investigated. Scanning electron microscopy (SEM) imaging showed homogeneously dispersed treated BN particles in the matrix. Furthermore, the thermal conductivity increased as the BN concentration was increased. This behavior was also observed when the filler size was increased. Experimentally obtained thermal conductivity values agree with the predicted data from the Maxwell-Eucken model well at less than 70 wt.% BN loading. A larger particle size BN-filled novolac resin exhibits a higher thermal conductivity than a smaller particle size BN-filled one. The combined use of 0.5 and 15 μm particles with a mass ratio of 2:1 achieved the maximum thermal conductivity among the testing systems. The thermal resistance properties of the composites were also studied.

Chemically stabilized epitaxial wurtzite-BN thin film

Science.gov (United States)

Vishal, Badri; Singh, Rajendra; Chaturvedi, Abhishek; Sharma, Ankit; Sreedhara, M. B.; Sahu, Rajib; Bhat, Usha; Ramamurty, Upadrasta; Datta, Ranjan

2018-03-01

We report on the chemically stabilized epitaxial w-BN thin film grown on c-plane sapphire by pulsed laser deposition under slow kinetic condition. Traces of no other allotropes such as cubic (c) or hexagonal (h) BN phases are present. Sapphire substrate plays a significant role in stabilizing the metastable w-BN from h-BN target under unusual PLD growth condition involving low temperature and pressure and is explained based on density functional theory calculation. The hardness and the elastic modulus of the w-BN film are 37 & 339 GPa, respectively measured by indentation along direction. The results are extremely promising in advancing the microelectronic and mechanical tooling industry.

Determination of prompt neutron decay constant of the AP-600 reactor core

International Nuclear Information System (INIS)

Surbakti, T.

1998-01-01

Determination of prompt neutron decay constant of the AP-600 reactor core has been performed using combination of two codes WIMS/D4 and Batan-2DIFF. The calculation was done at beginning of cycle and all of control rods pulled out. Cell generation from various kinds of core materials was done with 4 neutron energy group in 1-D transport code (WIMS/D4). The cell is considered for 1/4 fuel assembly in cluster model with square pitch arrange and then, the dimension of its unit cell is calculated. The unit cell consist of a fuel and moderator unit. The unit cell dimension as input data of WIMS/D4 code, called it annulus, is obtained from the equivalent unit cell. Macroscopic cross sections as output was used as input on neutron diffusion code Batan-2DIFF for core calculation as appropriate with three enrichment regions of the fuel of AP-600 core, namely 2, 2.5, and 3%. From result of diffusion code ( Batan-2DIFF) is obtained the value of delayed neutron fraction of 6.932E-03 and average prompt neutron life-time of 26.38 μs, so that the value of prompt neutron decay constant is 262.8 s-1. If it is compared the calculation result with the design value, the deviation are, for the design value of delayed neutron fraction is 7.5E-03, about 8% and the design value of average prompt neutron life time is 19.6 μs, about 34% respectively. The deviation because there are still unknown several core components of AP-600, so it didn't include in calculation yet

Engineering reliability in design phase: An application to AP-600 reactor passive safety system

International Nuclear Information System (INIS)

Majumdr, D.; Siahpush, A.S.; Hills, S.W.

1992-01-01

A computerized reliability enhancement methodology is described that can be used at the engineering design phase to help the designer achieve a desired reliability of the system. It can take into account the limitation imposed by a constraint such as budget, space, or weight. If the desired reliability of the system is known, it can determine the minimum reliabilities of the components, or how many redundant components are needed to achieve the desired reliability. This methodology is applied to examine the Automatic Depressurization System (ADS) of the new passively safe AP-600 reactor. The safety goal of a nuclear reactor dictates a certain reliability level of its components. It is found that a series parallel valve configuration instead of the parallel-series configuration of the four valves in one stage would improve the reliability of the ADS. Other valve characteristics and arrangements are explored to examine different reliability options for the system

Enhancement of surface mechanical properties by using TiN[BCN/BN] n/c-BN multilayer system

Science.gov (United States)

Moreno, H.; Caicedo, J. C.; Amaya, C.; Muñoz-Saldaña, J.; Yate, L.; Esteve, J.; Prieto, P.

2010-11-01

The aim of this work is to improve the mechanical properties of AISI 4140 steel substrates by using a TiN[BCN/BN] n/c-BN multilayer system as a protective coating. TiN[BCN/BN] n/c-BN multilayered coatings via reactive r.f. magnetron sputtering technique were grown, systematically varying the length period ( Λ) and the number of bilayers ( n) because one bilayer ( n = 1) represents two different layers ( tBCN + tBN), thus the total thickness of the coating and all other growth parameters were maintained constant. The coatings were characterized by Fourier transform infrared spectroscopy showing bands associated with h-BN bonds and c-BN stretching vibrations centered at 1400 cm -1 and 1100 cm -1, respectively. Coating composition and multilayer modulation were studied via secondary ion mass spectroscopy. Atomic force microscopy analysis revealed a reduction in grain size and roughness when the bilayer number ( n) increased and the bilayer period decreased. Finally, enhancement of mechanical properties was determined via nanoindentation measurements. The best behavior was obtained when the bilayer period ( Λ) was 80 nm ( n = 25), yielding the relative highest hardness (˜30 GPa) and elastic modulus (230 GPa). The values for the hardness and elastic modulus are 1.5 and 1.7 times greater than the coating with n = 1, respectively. The enhancement effects in multilayered coatings could be attributed to different mechanisms for layer formation with nanometric thickness due to the Hall-Petch effect; because this effect, originally used to explain increased hardness with decreasing grain size in bulk polycrystalline metals, has also been used to explain hardness enhancements in multilayered coatings taking into account the thickness reduction at individual single layers that make up the multilayered system. The Hall-Petch model based on dislocation motion within layered and across layer interfaces has been successfully applied to multilayered coatings to explain this

Enhancement of surface mechanical properties by using TiN[BCN/BN]n/c-BN multilayer system

International Nuclear Information System (INIS)

Moreno, H.; Caicedo, J.C.; Amaya, C.; Munoz-Saldana, J.; Yate, L.; Esteve, J.; Prieto, P.

2010-01-01

The aim of this work is to improve the mechanical properties of AISI 4140 steel substrates by using a TiN[BCN/BN] n /c-BN multilayer system as a protective coating. TiN[BCN/BN] n /c-BN multilayered coatings via reactive r.f. magnetron sputtering technique were grown, systematically varying the length period (Λ) and the number of bilayers (n) because one bilayer (n = 1) represents two different layers (t BCN + t BN ), thus the total thickness of the coating and all other growth parameters were maintained constant. The coatings were characterized by Fourier transform infrared spectroscopy showing bands associated with h-BN bonds and c-BN stretching vibrations centered at 1400 cm -1 and 1100 cm -1 , respectively. Coating composition and multilayer modulation were studied via secondary ion mass spectroscopy. Atomic force microscopy analysis revealed a reduction in grain size and roughness when the bilayer number (n) increased and the bilayer period decreased. Finally, enhancement of mechanical properties was determined via nanoindentation measurements. The best behavior was obtained when the bilayer period (Λ) was 80 nm (n = 25), yielding the relative highest hardness (∼30 GPa) and elastic modulus (230 GPa). The values for the hardness and elastic modulus are 1.5 and 1.7 times greater than the coating with n = 1, respectively. The enhancement effects in multilayered coatings could be attributed to different mechanisms for layer formation with nanometric thickness due to the Hall-Petch effect; because this effect, originally used to explain increased hardness with decreasing grain size in bulk polycrystalline metals, has also been used to explain hardness enhancements in multilayered coatings taking into account the thickness reduction at individual single layers that make up the multilayered system. The Hall-Petch model based on dislocation motion within layered and across layer interfaces has been successfully applied to multilayered coatings to explain this

Tunable localized surface plasmon resonances in one-dimensional h-BN/graphene/h-BN quantum-well structure

Science.gov (United States)

Kaibiao, Zhang; Hong, Zhang; Xinlu, Cheng

2016-03-01

The graphene/hexagonal boron-nitride (h-BN) hybrid structure has emerged to extend the performance of graphene-based devices. Here, we investigate the tunable plasmon in one-dimensional h-BN/graphene/h-BN quantum-well structures. The analysis of optical response and field enhancement demonstrates that these systems exhibit a distinct quantum confinement effect for the collective oscillations. The intensity and frequency of the plasmon can be controlled by the barrier width and electrical doping. Moreover, the electron doping and the hole doping lead to very different results due to the asymmetric energy band. This graphene/h-BN hybrid structure may pave the way for future optoelectronic devices. Project supported by the National Natural Science Foundation of China (Grant Nos. 11474207 and 11374217) and the Scientific Research Fund of Sichuan University of Science and Engineering, China (Grant No. 2014PY07).

Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

Energy Technology Data Exchange (ETDEWEB)

Groenwall, B; Ljungberg, L; Huebner, W; Stuart, W

1966-08-15

Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 {mu}g/cm{sup 2}). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in

Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

International Nuclear Information System (INIS)

Groenwall, B.; Ljungberg, L.; Huebner, W.; Stuart, W.

1966-08-01

Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 μg/cm 2 ). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in caustic

Account of requirements for modernization in VPBER-600 enhanced safety reactor instrumentation and control system development

International Nuclear Information System (INIS)

Shashkin, S.L.; Pobedonostsev, A.B.; Drumov, V.V.; Chudin, A.G.

1993-01-01

Nuclear power plant (NPP) with VPBER-600 reactor is a station of new generation. The specified term of reactor plant operation is 60 years and taking into account that the proposed term of starting the first power unit is on the turn of centuries one can definitely state that for Russia conditions VPBER-600 is a plant of 21 century. Such far removed term for NPP now in the stage of development as it can seem does not put the problems of modernization as first order tasks. But open-quotes...who does not think about future lives in the past.close quotes It is that the NPP instrumentation and control (I ampersand C) systems are in the most degree subjected to the influence of factors which favor their modifications. These factors can be arbitrarily divided into two groups: (1) inner factors, i.e. changes (failures, aging, etc) in I ampersand C components as well as changes dictated by technological reasons (change of equipment composition, control algorithms, operation modes); (2) outer factors, i.e. intensive development of information technologies and rapid improvement of electronic components. This presentation addresses the problem of modernization of the safety instrumentation for this next generation facility, and the research effort it will entail. The system is designed to allow for modernization, and the relatively easy adoption of new instrumentation and technology as it becomes available

Development of physical conceptions of fast reactors

International Nuclear Information System (INIS)

Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

2013-01-01

• Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 17, 2011, 6:00 AM status

International Nuclear Information System (INIS)

2011-01-01

This situation note is established according to the information gained on March 17, 2011, at 6:00 AM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

Criticality safety issues in the disposition of BN-350 spent fuel

International Nuclear Information System (INIS)

Schaefer, R. W.; Klann, R. T.; Koltyshev, S. M.; Krechetov, S.

2000-01-01

A criticality safety analysis has been performed as part of the BN-350 spent fuel disposition project being conducted jointly by the DOE and Kazakhstan. The Kazakhstan regulations are reasonably consistent with those of the DOE. The high enrichment and severe undermoderation of this fast reactor fuel has significant criticality safety consequences. A detailed modeling approach was used that showed some configurations to be safe that otherwise would be rejected. Reasonable requirements for design and operations were needed, and with them, all operations were found to be safe

Effects of fertile blanket on 600 MWth gas-cooled fast reactors: reactor and fuel cycle model

International Nuclear Information System (INIS)

Choi, Hang Bok

2002-07-01

A physics study has been performed to search for an optimum size of blanket for a 600 MWth gas-cooled fast reactor under fixed fuel and core specifications. The variables considered in this study are the reflector material, reflector thickness and blanket volume. The parametric calculations have shown that a positive breeding gain can be obtained by deploying 8 m 3 natural uranium blanket on the axial and radial boundaries of the core, surrounded by 40 cm Zr 3 Si 2 reflector. However the blanket core has disadvantages compared to the no-blanket core from the viewpoints of fuel fabrication cost and proliferation risk. On the other hand, the no-blanket core has large uncertainties in the possibility of achieving a positive breeding gain. Therefore further studies are recommended for the no-blanket option to improve the breeding gain and achieve a fissile self-sufficient fuel cycle, which is also proliferation-resistant. As an alternative, the blanket option can be considered, that ensures a positive breeding gain

Determination of filling-up time of the dousing system of CANDU-600 reactor

International Nuclear Information System (INIS)

Stefanescu, P.; Damian, C.

1997-01-01

The dousing system belongs to the envelope system being one of the fourth special safety system provided in the plant design in order to limit the release of the radioactive products to the population in case of serious damages. It is designed to reduce the top pressure and overpressure period within the envelope by taking over by the dousing water of the steam energy as resulted from an accident. Such an accident can be caused either by the loss of the primary cooling agent or the damage of the main steam pipe of the section situated in the envelope. At the same time this system is designed to achieve the function of maintaining sufficient water supply for the meeting of the dousing water necessities in the envelope and for the cooling of the reactor core in case of an accident with the loss of primary cooling agent. The paper has its target the calculation of the fill-up time of the dousing system of the CANDU-600 reactor, of the time period between the setting-up of the system and the moment the farthest nozzle starts dousing, respectively (the system douses at nominal flow within the envelope. (authors)

RELAP5/MOD3 AP600 problems

International Nuclear Information System (INIS)

Riemke, R.A.

1993-01-01

RELAP5/MOD3 is a reactor systems analysis code that has been developed jointly by the US Nuclear Regulatory Commission (USNRC) and a consortium consisting of several of the countries and domestic organizations that were members of the International Code Assessment and Applications Program (ICAP). The code is currently being used to simulate transients for the next generation of advanced light water reactors (ALWR's). One particular reactor design is the Westinghouse AP600 pressurized water reactor (PWR), which consists of two hot legs and four cold legs as well as passive emergency core cooling (ECC) systems. Initial calculations with RELAP5/MOD3 indicated that the code was not as robust as RELAP5/MOD2.5 with regard to AP600 calculations. Recent modifications in the areas of condensation wall heat transfer, interfacial heat transfer in the presence of noncondensibles, bubbly flow interfacial heat transfer, and time smoothing of both interfacial drag and interfacial heat transfer have improved the robustness, although more reliability is needed

Enhancement of surface mechanical properties by using TiN[BCN/BN]{sub n}/c-BN multilayer system

Energy Technology Data Exchange (ETDEWEB)

Moreno, H. [Laboratorio de Recubrimientos Duros, CDT-ASTIN SENA, Cali (Colombia); Caicedo, J.C., E-mail: Jcesarca@calima.univalle.edu.co [Grupo de Peliculas Delgadas, Universidad del Valle, Cali (Colombia); Amaya, C. [Grupo de Peliculas Delgadas, Universidad del Valle, Cali (Colombia); Munoz-Saldana, J. [Centro de Investigacion y de Estudios Avanzados del IPN, Unidad Queretaro, Mexico (Mexico); Yate, L.; Esteve, J. [Department de Fisica Aplicada i Optica, Universitat de Barcelona, Catalunya (Spain); Prieto, P. [Grupo de Peliculas Delgadas, Universidad del Valle, Cali (Colombia); Centro de Excelencia en Nuevos Materiales, CENM, Cali (Colombia)

2010-11-15

The aim of this work is to improve the mechanical properties of AISI 4140 steel substrates by using a TiN[BCN/BN]{sub n}/c-BN multilayer system as a protective coating. TiN[BCN/BN]{sub n}/c-BN multilayered coatings via reactive r.f. magnetron sputtering technique were grown, systematically varying the length period ({Lambda}) and the number of bilayers (n) because one bilayer (n = 1) represents two different layers (t{sub BCN} + t{sub BN}), thus the total thickness of the coating and all other growth parameters were maintained constant. The coatings were characterized by Fourier transform infrared spectroscopy showing bands associated with h-BN bonds and c-BN stretching vibrations centered at 1400 cm{sup -1} and 1100 cm{sup -1}, respectively. Coating composition and multilayer modulation were studied via secondary ion mass spectroscopy. Atomic force microscopy analysis revealed a reduction in grain size and roughness when the bilayer number (n) increased and the bilayer period decreased. Finally, enhancement of mechanical properties was determined via nanoindentation measurements. The best behavior was obtained when the bilayer period ({Lambda}) was 80 nm (n = 25), yielding the relative highest hardness ({approx}30 GPa) and elastic modulus (230 GPa). The values for the hardness and elastic modulus are 1.5 and 1.7 times greater than the coating with n = 1, respectively. The enhancement effects in multilayered coatings could be attributed to different mechanisms for layer formation with nanometric thickness due to the Hall-Petch effect; because this effect, originally used to explain increased hardness with decreasing grain size in bulk polycrystalline metals, has also been used to explain hardness enhancements in multilayered coatings taking into account the thickness reduction at individual single layers that make up the multilayered system. The Hall-Petch model based on dislocation motion within layered and across layer interfaces has been successfully applied to

Performance evaluation of control strategies for power maneuvering event of the KALIMER-600

International Nuclear Information System (INIS)

Seong, Seong-Hwan; Kim, Seong-O

2012-01-01

Highlights: ► The performance of three power control strategies of the KALIMER-600 was evaluated. ► There are turbine-, reactor- and feedwater-leading strategies in this study. ► For this, a performance analysis code was developed in this study. ► Simulation results show the turbine-leading is the best alternative. ► The feedwater-leading seems to be the second option. - Abstract: A sodium-cooled fast reactor named KALIMER-600 has been under development at KAERI. It is a pool-type reactor with the intermediate loops filled with sodium and has a superheated steam cycle with the once-through steam generators. Since the characteristic of the power control of the KALIMER-600 is expected to be different with that of a conventional power plant, the performance of the turbine-leading, reactor-leading and feedwater-leading control strategies for a power maneuvering event of the KALIMER-600 was evaluated in this study. The turbine-leading and reactor-leading strategies are very similar to those of a conventional water reactor but the feedwater-leading strategy is very similar to that of a fossil plant. Also, a performance analysis code which can analyze the plant dynamics of the KALIMER-600 and simulate the control actions during a power maneuvering event was developed. To evaluate the performance of control strategies, a simple power maneuvering event including a 10% step change and a ramp change with a rate of 5%/min was assumed and simulated. Through the simulation results, the turbine-leading strategy is proven to be very suitable for the KALIMER-600 and the feedwater-leading strategy for power maneuvering seems to be a good alternative for the power control. In further studies, various performance-related events such as the reactor power cutback, turbine runback and some transients will be evaluated and the best control strategy will be suggested.

Preliminary analyses of AP600 using RELAP5

International Nuclear Information System (INIS)

Modro, S.M.; Beelman, R.J.; Fisher, J.E.

1991-01-01

This paper presents results of preliminary analyses of the proposed Westinghouse Electric Corporation AP600 design. AP600 is a two loop, 600 MW (e) pressurized water reactor (PWR) arranged in a two hot leg, four cold leg nuclear steam supply system (NSSS) configuration. In contrast to the present generation of PWRs it is equipped with passive emergency core coolant (ECC) systems. Also, the containment and the safety systems of the AP600 interact with the reactor coolant system and each other in a more integral fashion than present day PWRs. The containment in this design is the ultimate heat sink for removal of decay heat to the environment. Idaho National Engineering Laboratory (INEL) has studied applicability of the RELAP5 code to AP600 safety analysis and has developed a model of the AP600 for the Nuclear Regulatory Commission. The model incorporates integral modeling of the containment, NSSS and passive safety systems. Best available preliminary design data were used. Nodalization sensitivity studies were conducted to gain experience in modeling of systems and conditions which are beyond the applicability of previously established RELAP5 modeling guidelines or experience. Exploratory analyses were then undertaken to investigate AP600 system response during postulated accident conditions. Four small break LOCA calculations and two large break LOCA calculations were conducted

Strain, stabilities and electronic properties of hexagonal BN bilayers

Science.gov (United States)

Fujimoto, Yoshitaka; Saito, Susumu

Hexagonal boron nitride (h-BN) atomic layers have been regarded as fascinating materials both scientifically and technologically due to the sizable band gap. This sizable band-gap nature of the h-BN atomic layers would provide not only new physical properties but also novel nano- and/or opto-electronics applications. Here, we study the first-principles density-functional study that clarifies the biaxial strain effects on the energetics and the electronic properties of h-BN bilayers. We show that the band gaps of the h-BN bilayers are tunable by applying strains. Furthermore, we show that the biaxial strains can produce a transition from indirect to direct band gaps of the h-BN bilayer. We also discuss that both AA and AB stacking patterns of h-BN bilayer become feasible structures because h-BN bilayers possess two different directions in the stacking patterns. Supported by MEXT Elements Strategy Initiative to Form Core Research Center through Tokodai Institute for Element Strategy, JSPS KAKENHI Grant Numbers JP26390062 and JP25107005.

Reactors. Nuclear propulsion ships

International Nuclear Information System (INIS)

Fribourg, Ch.

2001-01-01

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.)

Adsorption-induced gap states of h-BN on metal surfaces

Science.gov (United States)

Preobrajenski, A. B.; Krasnikov, S. A.; Vinogradov, A. S.; Ng, May Ling; Käämbre, T.; Cafolla, A. A.; Mårtensson, N.

2008-02-01

The formation of hexagonal boron nitride (h-BN) monolayers on Ni(111), Rh(111), and Pt(111) has been studied by a combination of x-ray emission, angle-resolved valence band photoemission, and x-ray absorption in search for interface-induced gap states of h-BN . A significant density of both occupied and unoccupied gap states with N2p and B2p characters is observed for h-BN/Ni(111) , somewhat less for h-BN/Rh(111) and still less for h-BN/Pt(111) . X-ray emission shows that the h-BN monolayer is chemisorbed strongly on Ni(111) and very weakly on Pt(111). We associate the gap states of h-BN adsorbed on the transition metal surfaces with the orbital mixing and electron sharing at the interface because their density increases with the growing strength of chemisorption.

How lithium atoms affect the first hyperpolarizability of BN edge-doped graphene.

Science.gov (United States)

Song, Yao-Dong; Wu, Li-Ming; Chen, Qiao-Ling; Liu, Fa-Kun; Tang, Xiao-Wen

2016-01-01

How do lithium atoms affect the first hyperpolarizability (β0) of boron-nitrogen (BN) edge-doped graphene. In this work, using pentacene as graphene model, Lin@BN-1 edge-doped pentacene and Lin@BN-2 edge-doped pentacene (n = 1, 5) were designed to study this problem. First, two models (BN-1 edge-doped pentacene, and BN-2 edge-doped pentacene ) were formed by doping the BN into the pentacene with different order, and then Li@BN-1 edge-doped pentacene and Li@ BN-2 edge-doped pentacene were obtained by substituting the H atom in BN edge-doped pentacene with a Li atom. The results show that the first hyperpolarizabilities of BN-1 edge-doped pentacene and Li@BN-1 edge-doped pentacene were 4059 a.u. and 6249 a.u., respectively; the first hyperpolarizabilities of BN-2 edge-doped pentacene and Li@BN-2 edge-doped pentacene were 2491 a.u. and 4265 a.u., respectively. The results indicate that the effect of Li substitution is to greatly increase the β0 value. To further enhance the first hyperpolarizability, Li5@ BN-1 edge-doped pentacene and Li5@BN-2 edge-doped pentacene were designed, and were found to exhibit considerably larger first hyperpolarizabilities (β0) (12,112 a.u. and 7921a.u., respectively). This work may inspire further study of the nonlinear properties of BN edge-doped graphene.

Correlation of yield stress and microhardness in 08Cr16Ni11Mo3 stainless steel irradiated to high dose in the BN-350 fast reactor

International Nuclear Information System (INIS)

Gusev, M.N.; Maksimkin, O.P.; Tivanova, O.V.; Silnaygina, N.S.; Garner, F.A.

2006-01-01

The relationship between the microhardness and the engineering yield stress in 08Cr16Ni11Mo3 steel after irradiation in the BN-350 reactor has been experimentally derived and agrees with a previously published correlation developed by Toloczko for unirradiated 316 in a variety of cold-work conditions. Even more importantly, when the correlation is derived in the K Δ format where the correlation involves changes in the two properties, excellent agreement is found with a universal K Δ correlation developed by Busby and coworkers. Additionally, this report points out that microhardness measurements must take into account that sodium exposure at high temperature and neutron fluence alters the metal surface to produce ferrite, and therefore the altered layers should be removed prior to testing

MOX fuel fabrication, in reactor performance and improvement

International Nuclear Information System (INIS)

Vliet, J. van; Deramaix, P.; Nigon, J.L.; Fournier, W.

1998-01-01

In Europe, MOX fuel for light water reactors (LWRs) has first been manufactured in Belgium and Germany. Belgonucleaire (BN) loaded the first MOX assembly in the BR3 Pressurised Water Reactor (PWR) in 1963. In June 1998, more than 750 tHM LWR MOX fuel assemblies were manufactured on a industrial scale in Europe without any particular difficulty relating to fuel fabrication, reactor operation or fuel behaviour. So, today plutonium recycling through MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants. In this field, COGEMA and BELGONUCLEAIRE are the main actors by operating simultaneously three complete multidesign fuel production plants: MELOX plant (in Marcoule), CADARACHE plant and P0 plant (in Dessel, Belgium). Present MOX production capacity available to COGEMA and BN fits 175 tHM per year and is to be extended to reach about 325 tHM in the year 2000. This will represent 75% of the total MOX fabrication capacity in Europe. The industrial mastery and the high production level in MOX fabrication assured by high technology processes confer to these companies a large expertise for Pu recycling. This allows COGEMA and BN to be major actors in Pu-based fuels in the coming second nuclear era with advanced fuel cycles. (author)

ALARA radiation considerations for the AP600 reactor

Energy Technology Data Exchange (ETDEWEB)

Lau, F.L. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-03-01

The radiation design of the AP600 reactor plant is based on an average annual occupational radiation exposure (ORE) of 100 man-rem. As a design goal we have established a lower value of 70 man-rem per year. And, with our current design process, we expect to achieve annual exposures which are well below this goal. To accomplish our goal we have established a process that provides criteria, guidelines and customer involvement to achieve the desired result. The criteria and guidelines provide the shield designer, as well as the systems and plant layout designers with information that will lead to an integrated plant design that minimizes personnel exposure and yet is not burdened with complicated shielding or unnecessary component access limitations. Customer involvement is provided in the form of utility input, design reviews and information exchange. Cooperative programs with utilities in the development of specific systems or processes also provides for an ALARA design. The results are features which include ALARA radiation considerations as an integral part of the plant design and a lower plant ORE. It is anticipated that a further reduction in plant personnel exposures will result through good radiological practices by the plant operators. The information in place to support and direct the plant designers includes the Utility Requirements Document (URD), Federal Regulations, ALARA guidelines, radiation design information and radiation and shielding design criteria. This information, along with the utility input, design reviews and information feedback, will contribute to the reduction of plant radiation exposure levels such that they will be less than the stated goals.

Irradiation Creep and Swelling of Russian Ferritic-Martensitic Steels Irradiated to Very High Exposures in the BN-350 Fast Reactor at 305-335 degrees C

International Nuclear Information System (INIS)

Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

2003-01-01

Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb-Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

Characterizing and packaging BN-350 spent fuel for long-term dry storage

International Nuclear Information System (INIS)

Lambert, J. D. B.; Bolshinsky, I.; Haues, S.L.; Allen, K.J.; Howden, E.A.; Hill, R.N.; Planchon, H.P.; Staples, P.; Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.K.; Maev, V.; Dumchev, I. A.

2000-01-01

The Republic of Kazakhstan is being assisted by the U.S. Department of Energy in preparing spent fuel from the BN-350 fast reactor for long term dry storage. Argonne National Laboratory was assigned responsibility for the physical and nuclear characterization of the spent fuel, for the design and safety analysis of 6-pac and 4-pac canisters used to contain spent fuel assemblies for storage, and for the design, testing and installation of a closure station at the reactor in which the canisters of fuel are dried, filled with inert gas and welded shut. This paper briefly describes the specialized components and equipment used, the process followed, and experience gained in packaging the spent fuel. Olsen et al and Schaefer separately discuss overall safety and criticality considerations of the packaging process in parallel papers to this conference

Effects of existing evaluated nuclear data files on neutronics characteristics of the BFS-62-3A critical assembly benchmark model

International Nuclear Information System (INIS)

Semenov, Mikhail

2002-11-01

This report is continuation of studying of the experiments performed on BFS-62-3A critical assembly in Russia. The objective of work is definition of the cross section uncertainties on reactor neutronics parameters as applied to the hybrid core of the BN-600 reactor of Beloyarskaya NPP. Two-dimensional benchmark model of BFS-62-3A was created specially for these purposes and experimental values were reduced to it. Benchmark characteristics for this assembly are 1) criticality; 2) central fission rate ratios (spectral indices); and 3) fission rate distributions in stainless steel reflector. The effects of nuclear data libraries have been studied by comparing the results calculated using available modern data libraries - ENDF/B-V, ENDF/B-VI, ENDF/B-VI-PT, JENDL-3.2 and ABBN-93. All results were computed by Monte Carlo method with the continuous energy cross-sections. The checking of the cross sections of major isotopes on wide benchmark criticality collection was made. It was shown that ENDF/B-V data underestimate the criticality of fast reactor systems up to 2% Δk. As for the rest data, the difference between each other in criticality for BFS-62-3A is around 0.6% Δk. However, taking into account the results obtained for other fast reactor benchmarks (and steel-reflected also), it may conclude that the difference in criticality calculation results can achieve 1% Δk. This value is in a good agreement with cross section uncertainty evaluated for BN-600 hybrid core (±0.6% Δk). This work is related to the JNC-IPPE Collaboration on Experimental Investigation of Excess Weapons Grade Pu Disposition in BN-600 Reactor Using BFS-2 Facility. (author)

Safety analysis for key design features of KALIMER-600 design concept

International Nuclear Information System (INIS)

Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S.

2005-03-01

KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed

Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

International Nuclear Information System (INIS)

Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

1996-01-01

Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

Preparations and thermal properties of micro- and nano-BN dispersed HDPE composites

International Nuclear Information System (INIS)

Jung, Jinwoo; Kim, Jaewoo; Uhm, Young Rang; Jeon, Jae-Kyun; Lee, Sol; Lee, Hi Min; Rhee, Chang Kyu

2010-01-01

The thermal properties of micro-sized boron nitride (BN) and nano-sized BN dispersed high density polyethylene (HDPE) composites were investigated by means of differential scanning calorimetry (DSC) and thermo-gravimetric analysis (TGA). Nano-BN powder was prepared by using a ball mill process before it was mixed in HDPE. To enhance the dispersivity of nano-BN in the polymer matrix, the surfaces of the nano-particles were treated with low density polyethylene (LDPE) which was dissolved in the cyclohexane solvent. The average particle sizes of micro-BN powder and LDPE coated nano-BN powder were ∼10 μm and ∼100 nm respectively. Dispersion and distribution of 5 wt% and 20 wt% of micro-BN and nano-BN respectively mixed in HDPE were observed by using the scanning electron microscope (SEM). According to the thermal analyses of pure HDPE, micro-BN/HDPE, and nano-BN/HDPE, 20 wt% nano-BN/HDPE composite shows the lowest enthalpy of fusion (ΔH m ) and better thermal conductive characteristics compared to the others.

Development of Preliminary PIRTs of Thermal-Hydraulic Phenomena for KALIMER-600

International Nuclear Information System (INIS)

Kwon, Young Min; Jeong, Hae Yong; Ha, Kwi Seok; Chang, Won Pyo

2009-01-01

Sodium Cooled Fast Reactors (SFRs) are the most technologically developed of the GEN IV systems. The primary mission of the SFRs is the management of high-level wastes, in particular management of plutonium and other actinides. The SFR system is the nearest-term actinide management system among the GEN-IV system candidates. The mission of the SFR can be extended to electricity production if design innovations that reduce capital cost. KAERI has been performing design studies of KALIMER-600 at the conceptual level. To bring KALIMER-600 to deployment, several technology gaps in fuel cycle and reactor system must be closed. Research on both sides of the fuel cycle and the reactor system is necessary to bring KALIMER-600 to deployment. For the reactor system, technology gaps exist in assurance or verification of passive safety, and completion of the metallic fuel database including irradiation performance data. R and D programs for the KALIMER-600 safety are necessary to support the SFR deployment. The safety R and D challenges for the KALIMER-600 in the context of the GEN IV systems are: (a) to verify the predictability and effectiveness of the inherent passive benign responses to design basis events and accommodated beyond design basis events (b) to provide assurance that accommodated beyond design basis events considered in licensing can be sustained without loss of coolability of fuel and structural integrity. The Phenomena Identification and Ranking Table (PIRT) is an effective tool for providing an expert assessment of safety-related phenomena and for assessing R and D needs for KALIMER-600 licensing. The nine-step PIRT process has been established as a methodology for providing expert assessments of safety-relevant phenomena

Electronic structure of graphene- and BN-supported phosphorene

Science.gov (United States)

Davletshin, Artur R.; Ustiuzhanina, Svetlana V.; Kistanov, Andrey A.; Saadatmand, Danial; Dmitriev, Sergey V.; Zhou, Kun; Korznikova, Elena A.

2018-04-01

By using first-principles calculations, the effects of graphene and boron nitride (BN) substrates on the electronic properties of phosphorene are studied. Graphene-supported phosphorene is found to be metallic, while the BN-supported phosphorene is a semiconductor with a moderate band gap of 1.02 eV. Furthermore, the effects of the van der Waals interactions between the phosphorene and graphene or BN layers by means of the interlayer distance change are investigated. It is shown that the interlayer distance change leads to significant band gap size modulations and direct-indirect band gap transitions in the phosphorene-BN heterostructure. The presented band gap engineering of phosphorene may be a powerful technique for the fabrication of high-performance phosphorene-based nanodevices.

Tunable magnetotransport in Fe/hBN/graphene/hBN/Pt(Fe) epitaxial multilayers

Science.gov (United States)

Magnus Ukpong, Aniekan

2018-03-01

Theoretical and computational analysis of the magnetotransport properties and spin-transfer torque field-induced switching of magnetization density in vertically-stacked multilayers is presented. Using epitaxially-capped free layers of Pt and Fe, atom-resolved magnetic moments and spin-transfer torques are computed at finite bias. The calculations are performed within linear response approximation to the spin-density reformulation of the van der Waals density functional theory. Dynamical spin excitations are computed as a function of a spin-transfer torque induced magnetic field along the magnetic easy axis, and the corresponding spin polarization perpendicular to the easy axis is obtained. Bias-dependent giant anisotropic magnetoresistance of up to 3200% is obtained in the nonmagnetic-metal-capped Fe/hBN/graphene/hBN/Pt multilayer architecture. Since this specific heterostructure is not yet fabricated and characterized, the predicted high performance has not been demonstrated experimentally. Nevertheless, similar calculations performed on the Fe/hBN/Co stack show that the tunneling magnetoresistance obtained at the Fermi-level is in excellent agreement with results of recent magnetotransport measurements on magnetic tunnel junctions that contain the monolayer hBN tunnel region. The magnitude of the spin-transfer torque is found to increase as the tunneling spin current increases, and this activates the magnetization switching process due to increased charge accumulation. This mechanism causes substantial spin backflow, which manifests as rapid undulations in the bias-dependent tunneling spin currents. The implication of these findings on the design of nanoscale spintronic devices with spin-transfer torque tunable magnetization density is discussed. Insights derived from this study are expected to enhance the prospects for developing and integrating artificially assembled van der Waals multilayer heterostructures as the preferred material platform for efficient

Bias induced up to 100% spin-injection and detection polarizations in ferromagnet/bilayer-hBN/graphene/hBN heterostructures

NARCIS (Netherlands)

Gurram, Mallikarjuna; Omar, Siddharta; van Wees, Bart

2017-01-01

We study spin transport in a fully hBN encapsulated monolayer-graphene van der Waals heterostructure at room temperature. A top-layer of bilayer-hBN is used as a tunnel barrier for spin-injection and detection in graphene with ferromagnetic cobalt electrodes. We report surprisingly large and

Side-gate modulation effects on high-quality BN-Graphene-BN nanoribbon capacitors

International Nuclear Information System (INIS)

Wang, Yang; Chen, Xiaolong; Ye, Weiguang; Wu, Zefei; Han, Yu; Han, Tianyi; He, Yuheng; Cai, Yuan; Wang, Ning

2014-01-01

High-quality BN-Graphene-BN nanoribbon capacitors with double side-gates of graphene have been experimentally realized. The double side-gates can effectively modulate the electronic properties of graphene nanoribbon capacitors. By applying anti-symmetric side-gate voltages, we observed significant upward shifting and flattening of the V-shaped capacitance curve near the charge neutrality point. Symmetric side-gate voltages, however, only resulted in tilted upward shifting along the opposite direction of applied gate voltages. These modulation effects followed the behavior of graphene nanoribbons predicted theoretically for metallic side-gate modulation. The negative quantum capacitance phenomenon predicted by numerical simulations for graphene nanoribbons modulated by graphene side-gates was not observed, possibly due to the weakened interactions between the graphene nanoribbon and side-gate electrodes caused by the Ga + beam etching process

Resistance to BN myelogenous leukemia in rat radiation chimeras

International Nuclear Information System (INIS)

Singer, D.E.; Haynor, D.R.; Williams, R.M

1980-01-01

Lewis → LBNFl rat radiation chimeras showed marked resistance to transplanted BN myelogenous leukemia when compared to naive LBNFl, LBNFl → LBNFl, or BN → LBNFl. This occurred in the absence of overt graft versus host disease or of anti-BN response in mixed lymphocyte culture. Bone marrow specific antigens may serve as the target of the resistance mechanism. (author)

Nanoindentation of ultra-hard cBN films: A molecular dynamics study

Energy Technology Data Exchange (ETDEWEB)

Huang, Cheng [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); Peng, Xianghe, E-mail: xhpeng@cqu.edu.cn [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); State Key Laboratory of Coal Mine Disaster Dynamics and Control, Chongqing University, Chongqing 400044 (China); Chongqing Key Laboratory of Heterogeneous Material Mechanics, Chongqing University, Chongqing 400044 (China); Fu, Tao, E-mail: futaocqu@163.com [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); Zhao, Yinbo; Feng, Chao; Lin, Zijun [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); Li, Qibin [College of Aerospace Engineering, Chongqing University, Chongqing 400044 (China); Chongqing Key Laboratory of Heterogeneous Material Mechanics, Chongqing University, Chongqing 400044 (China)

2017-01-15

Highlights: • We optimize tersoff potential to better simulate the BN. • We perform respectively the nanoindentations on the (001) and (111) surface of cBN. • The main slip system of cBN under nanoindentation is {111}<110>. • Temperature has a significant effect on the mechanical properties of cBN. - Abstract: Cubic Boron nitride (cBN) exhibits excellent mechanical properties including high strength, hardness and thermal resistance, etc. We optimized the parameters in the Tersoff interatomic potential for cBN based on its cohesive energy, lattice parameter, elastic constants, surface energy and stacking fault energy. We performed with molecular dynamics (MD) simulations the nanoindentation on the (001) and (111) surface of monocrystalline cBN thin films to study the deformation mechanisms and the effects of temperature and substrate orientation. It was found that during the indentation plastic deformation is mainly stress-induced slips of dislocations along {111}<110> orientations. It was also found that the hardness of cBN depends strongly on temperature, and the capability of plastic deformation is enhanced with the increase of temperature.

Friction and wear behavior of laser cladding Ni/hBN self-lubricating composite coating

International Nuclear Information System (INIS)

Zhang Shitang; Zhou Jiansong; Guo Baogang; Zhou Huidi; Pu Yuping; Chen Jianmin

2008-01-01

Ni/hBN coating was successfully prepared on 1Cr18Ni9Ti stainless steel substrate by means of laser cladding. The microhardness profile of the composite coating along the depth direction was measured, while its cross-sectional microstructures and phase compositions were analyzed by means of scanning electron microscopy and X-ray diffraction. Moreover, the friction and wear behavior of the composite coatings sliding against Si 3 N 4 from ambient to 800 deg. C was evaluated using a ball-on-disc friction and wear tester, and the worn surface morphologies of the composite coatings and counterpart ceramic balls were observed using a scanning electron microscope. At the same time, the worn surfaces of the ceramic balls were also analyzed using a 3D non-contact surface mapping profiler as well. It was found that the laser cladding Ni/hBN coating on the stainless steel substrate had high microhardness and good friction-reducing and antiwear abilities at elevated temperatures up to 800 deg. C. The composite coating registered slightly increased friction coefficient and wear rate as the temperature rose from ambient to 100 deg. C; then the friction coefficient and wear rate decreased with increasing temperature up to 800 deg. C (with the slight increase in the wear rate at 700 deg. C and 800 deg. C to be an exception). The laser cladding Ni/hBN coating was dominated by mixed adhesion and abrasive wear as it slid against the ceramic ball below 300 deg. C. With further increase in the test temperature up to 400 deg. C and above, it was characterized by mild adhesion wear and plastic deformation. Since the laser cladding Ni/hBN coating registered an increased wear rate at temperatures of 600 deg. C and above, it was not suggested to be used for wear prevention and protection of the stainless steel at elevated temperature above 800 deg. C

Friction and wear behavior of laser cladding Ni/hBN self-lubricating composite coating

Energy Technology Data Exchange (ETDEWEB)

Zhang Shitang [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Graduate School, Chinese Academy of Sciences, Beijing 100039 (China); Zhou Jiansong [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Guo Baogang [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Graduate School, Chinese Academy of Sciences, Beijing 100039 (China); Zhou Huidi [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Pu Yuping [Central Iron and Steel Research Institute, Beijing 100081 (China); Chen Jianmin [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China)], E-mail: chenjm@lzb.ac.cn

2008-09-15

Ni/hBN coating was successfully prepared on 1Cr18Ni9Ti stainless steel substrate by means of laser cladding. The microhardness profile of the composite coating along the depth direction was measured, while its cross-sectional microstructures and phase compositions were analyzed by means of scanning electron microscopy and X-ray diffraction. Moreover, the friction and wear behavior of the composite coatings sliding against Si{sub 3}N{sub 4} from ambient to 800 deg. C was evaluated using a ball-on-disc friction and wear tester, and the worn surface morphologies of the composite coatings and counterpart ceramic balls were observed using a scanning electron microscope. At the same time, the worn surfaces of the ceramic balls were also analyzed using a 3D non-contact surface mapping profiler as well. It was found that the laser cladding Ni/hBN coating on the stainless steel substrate had high microhardness and good friction-reducing and antiwear abilities at elevated temperatures up to 800 deg. C. The composite coating registered slightly increased friction coefficient and wear rate as the temperature rose from ambient to 100 deg. C; then the friction coefficient and wear rate decreased with increasing temperature up to 800 deg. C (with the slight increase in the wear rate at 700 deg. C and 800 deg. C to be an exception). The laser cladding Ni/hBN coating was dominated by mixed adhesion and abrasive wear as it slid against the ceramic ball below 300 deg. C. With further increase in the test temperature up to 400 deg. C and above, it was characterized by mild adhesion wear and plastic deformation. Since the laser cladding Ni/hBN coating registered an increased wear rate at temperatures of 600 deg. C and above, it was not suggested to be used for wear prevention and protection of the stainless steel at elevated temperature above 800 deg. C.

Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

2015-02-15

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Directory of Open Access Journals (Sweden)

Yong Kyo Lee

2015-02-01

Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Calculated carrier mobility of h-BN/γ-InSe/h-BN van der Waals heterostructures

Science.gov (United States)

Kang, P.; Michaud-Rioux, V.; Kong, X.-H.; Yu, G.-H.; Guo, H.

2017-12-01

Recent experiments reported excellent transport properties of two-dimensional (2D) van der Waals (vdW) heterostructures made of atomically thin InSe layers encapsulated by two hBN capping layers (ISBN). The carrier mobility of the ISBN films exceeded μ ˜ 1.2× {{10}4} \\text{c}{{\\text{m}}2} {{\\text{V}}-1} {{\\text{s}}-1} at low temperature, much higher than that of pristine InSe films. It has been puzzling why the relatively inert hBN capping layer could so drastically enhance mobility of the ISBN composite. Using a state-of-the-art first principles method, we have calculated phonon limited carrier mobility of 18 different ISBN films and 6 pristine InSe films with different thicknesses, the largest system containing 2212 atoms. The hBN capping layer significantly alters the elastic stiffness coefficient as compared with pure InSe—thus the acoustic phonons in the ISBN composite—giving rise to the observed large mobility of ISBN films. Of the 18 calculated ISBN films, the ones with no strain at the hBN/InSe interface possess the highest electron mobility, reaching 4340~\\text{c}{{\\text{m}}2}~{{\\text{V}}-1}~{{\\text{s}}-1} at room temperature, which could easily go over {{10}4}~\\text{c}{{\\text{m}}2}~{{\\text{V}}-1}~{{\\text{s}}-1} at low temperatures. We conclude that the mechanical properties of the composite 2D vdW ISBN material play the crucial role for inducing the large carrier mobility, a principle that could be applied to many other 2D vdW heterostructures.

Electronic Properties of Curved and Defective 2-D BN Nanostructures

Science.gov (United States)

Beach, Kory; Terrones, Humberto; Raeliarijaona, Aldo; Siegel, Ross; Florio, Fred

Density functional theory (DFT) with local density approximation (LDA) pseudopotentials is used to calculate the band structure and density of states of various novel 2-D BN nanostructures. Three types of systems are studied: Schwarzites, a Haeckelite, and an h-BN monolayer. Schwarzites are negatively curved structures in which the curvature is due to the introduction of octagonal rings of alternating boron and nitrogen atoms. In particular, three families of Schwarzites are analyzed: P, G and IWP. The Haeckelites on the other hand, are flat layers composed of squares and octagons of BN. It is found that all these BN allotropes are metastable in which the band gap is direct and smaller than the most stable system, h-BN. National Science Foundation (EFRI-1433311).

Running-in strategies for the low-enriched 600 MW(e) D-HHT reactor. Part 1. Comparison of different on-load refuelling schemes

Energy Technology Data Exchange (ETDEWEB)

Hansen, U

1973-03-14

This paper presents detailed burn-up calculations and fuel management strategies for the Dragon-HHT, D-HHT, reference core. The reference layout was chosen from the outcome of a design survey with the 1-D equilibrium fuel cycle code FLATTER. The decision was based on aspects of engineering and economics. The purpose of the investigation is to devise a suitable first core, follow the irradiation history of the fuel and the general behaviour of the reactor during the first core replacements until equilibrium operating conditions are reached. A detailed description of time dependant burn-up and spatial power production for specified reactivity limits is required. For this purpose the reactor code system VSOP was employed. Different combinations of the parameters are investigated and the influence on reactor operation and economics discussed. From the strategy analysis a reference fuel management scheme is chosen for the low enriched 600 MW(e) D-HHT reactor.

Investigation on two abnormal phenomena about thermal conductivity enhancement of BN/EG nanofluids.

Science.gov (United States)

Li, Yanjiao; Zhou, Jing'en; Luo, Zhifeng; Tung, Simon; Schneider, Eric; Wu, Jiangtao; Li, Xiaojing

2011-07-09

The thermal conductivity of boron nitride/ethylene glycol (BN/EG) nanofluids was investigated by transient hot-wire method and two abnormal phenomena was reported. One is the abnormal higher thermal conductivity enhancement for BN/EG nanofluids at very low-volume fraction of particles, and the other is the thermal conductivity enhancement of BN/EG nanofluids synthesized with large BN nanoparticles (140 nm) which is higher than that synthesized with small BN nanoparticles (70 nm). The chain-like loose aggregation of nanoparticles is responsible for the abnormal increment of thermal conductivity enhancement for the BN/EG nanofluids at very low particles volume fraction. And the difference in specific surface area and aspect ratio of BN nanoparticles may be the main reasons for the abnormal difference between thermal conductivity enhancements for BN/EG nanofluids prepared with 140- and 70-nm BN nanoparticles, respectively.

Investigation on two abnormal phenomena about thermal conductivity enhancement of BN/EG nanofluids

Directory of Open Access Journals (Sweden)

Wu Jiangtao

2011-01-01

Full Text Available Abstract The thermal conductivity of boron nitride/ethylene glycol (BN/EG nanofluids was investigated by transient hot-wire method and two abnormal phenomena was reported. One is the abnormal higher thermal conductivity enhancement for BN/EG nanofluids at very low-volume fraction of particles, and the other is the thermal conductivity enhancement of BN/EG nanofluids synthesized with large BN nanoparticles (140 nm which is higher than that synthesized with small BN nanoparticles (70 nm. The chain-like loose aggregation of nanoparticles is responsible for the abnormal increment of thermal conductivity enhancement for the BN/EG nanofluids at very low particles volume fraction. And the difference in specific surface area and aspect ratio of BN nanoparticles may be the main reasons for the abnormal difference between thermal conductivity enhancements for BN/EG nanofluids prepared with 140- and 70-nm BN nanoparticles, respectively.

Purification and characterization of the bacteriocin Thuricin Bn1 produced by Bacillus thuringiensis subsp. kurstaki Bn1 isolated from a hazelnut pest.

Science.gov (United States)

Ugras, Serpil; Sezen, Kazim; Kati, Hatice; Demirbag, Zihni

2013-02-01

A novel bioactive molecule produced by Bacillus thuringiensis subsp. kurstaki Bn1 (Bt-Bn1), isolated from a common pest of hazelnut, Balaninus nucum L. (Coleoptera: Curculionidae), was determined, purified, and characterized in this study. The Bt-Bn1 strain was investigated for antibacterial activity with an agar spot assay and well diffusion assay against B. cereus, B. weinhenstephenensis, L. monocytogenes, P. savastanoi, P. syringae, P. lemoignei, and many other B. thuringiensis strains. The production of bioactive molecule was determined at the early logarithmic phase in the growth cycle of strain Bt-Bn1 and its production continued until the beginning of the stationary phase. The mode of action of this molecule displayed bacteriocidal or bacteriolytic effect depending on the concentration. The bioactive molecule was purified 78-fold from the bacteria supernatant with ammonium sulfate precipitation, dialysis, ultrafiltration, gel filtration chromatography, and HPLC, respectively. The molecular mass of this molecule was estimated via SDS-PAGE and confirmed by the ESI-TOFMS as 3,139 Da. The bioactive molecule was also determined to be a heat-stable, pH-stable (range 6-8), and proteinase K sensitive antibacterial peptide, similar to bacteriocins. Based on all characteristics determined in this study, the purified bacteriocin was named as thuricin Bn1 because of the similarities to the previously identified thuricin-like bacteriocin produced by the various B. thuringiensis strains. Plasmid elution studies showed that gene responsible for the production of thuricin Bn1 is located on the chromosome of Bt-Bn1. Therefore, it is a novel bacteriocin and the first recorded one produced by an insect originated bacterium. It has potential usage for the control of many different pathogenic and spoilage bacteria in the food industry, agriculture, and various other areas.

The effect of incorporated self-lubricated BN(h) particles on the tribological properties of Ni–P/BN(h) composite coatings

Energy Technology Data Exchange (ETDEWEB)

Hsu, Chih-I., E-mail: s1322509@gmail.com [School of Defense Science, Chung Cheng Institute of Technology, National Defense University, Taoyuan, Taiwan (China); Hou, Kung-Hsu, E-mail: khou@ndu.edu.tw [Department of Power Vehicle and Systems Engineering, Chung Cheng Institute of Technology, National Defense University, Taoyuan, Taiwan (China); Ger, Ming-Der, E-mail: mingderger@gmail.com [Department of Chemistry and Material Engineering, Chung Cheng Institute of Technology, National Defense University, Taoyuan, Taiwan (China); Wang, Gao-Liang, E-mail: wanggl@takming.edu.tw [Department of Marketing Management, Takming University of Science and Technology, Taipei, Taiwan (China)

2015-12-01

Highlights: • The Ni-P-BN(h) coatings were prepared by electroless plating techniques in this research. • Surfactant CTAB resulting in a uniform dispersion of particles in Ni-P coating. • CTAB with a positive effect on the tribological performance of Ni–P/BN(h) coatings. • Frictional tests results show that optimal friction coefficient would be decreased 75%. • Wear resistance of the Ni-P/BN(h) coating is higher about 10 times Ni–P coatings. - Abstract: Ni–P/BN(h) composite coatings are prepared by means of the conventional electroless plating from the bath containing up to 10.0 g/l of hexagonal boron nitride particles with size 0.5 μm. The Ni–P coating is also prepared as a comparison. Cationic surfactant cetyltrimethylammonium bromide (CTAB) is used to stabilize the electrolyte, and the optimum CTAB concentration resulting in a nonagglomerated dispersion of particles is obtained using a dispersion stability analyzer. Morphology of the coatings and the effect of incorporated particles on coating structure and composition are investigated via scanning electron microscopy, field emission electron probe micro-analyzer and X-ray diffraction analysis. Hardness, roughness, friction coefficient and wear resistance of the coatings are also evaluated using Vickers microhardness tester, atomic force microscopy and ball-on disk machine. The presence of CTAB in the depositing bath has a positive effect on the surface roughness and performance of Ni–P/BN(h) composite coatings. The friction and wear tests results show that incorporation of 14.5 vol% BN(h) particles into the Ni–P coating lowers the coating friction coefficient by about 75% and the wear resistance of the Ni–P composites is approximately 10 times higher than Ni–P coating.

Near-field heat transfer between graphene/hBN multilayers

OpenAIRE

Zhao, Bo; Guizal, Brahim; Zhang, Zhuomin M.; Fan, Shanhui; Antezza, Mauro

2017-01-01

We study the radiative heat transfer between multilayer structures made by a periodic repetition of a graphene sheet and a hexagonal boron nitride (hBN) slab. Surface plasmons in a monolayer graphene can couple with a hyperbolic phonon polaritons in a single hBN film to form hybrid polaritons that can assist photon tunneling. For periodic multilayer graphene/hBN structures, the stacked metallic/dielectric array can give rise to a further effective hyperbolic behavior, in addition to the intri...

On the Stability of c-BN-Reinforcing Particles in Ceramic Matrix Materials

Directory of Open Access Journals (Sweden)

Anne-Kathrin Wolfrum

2018-02-01

Full Text Available Cubic boron nitride (c-BN composites produced at high pressures and temperatures are widely used as cutting tool materials. The advent of new, effective pressure-assisted densification methods, such as spark plasma sintering (SPS, has stimulated attempts to produce these composites at low pressures. Under low-pressure conditions, however, transformation of c-BN to the soft hexagonal BN (h-BN phase can occur, with a strong deterioration in hardness and wear. In the present work, the influence of secondary phases (B2O3, Si3N4, and oxide glasses on the transformation of c-BN was studied in the temperature range between 1100 °C and 1575 °C. The different heat treated c-BN particles and c-BN composites were analyzed by SEM, X-ray diffraction, and Raman spectroscopy. The transformation mechanism was found to be kinetically controlled solution–diffusion–precipitation. Given a sufficiently low liquid phase viscosity, the transformation could be observed at temperatures as low as 1200 °C for the c-BN–glass composites. In contrast, no transformation was found at temperatures up to 1575 °C when no liquid oxide phase is present in the composite. The results were compared with previous studies concerning the c-BN stability and the c-BN phase diagram.

Development of technology for next generation reactor - Development of next generation reactor in Korea -

Energy Technology Data Exchange (ETDEWEB)

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

1993-09-01

The project, development of next generation reactor, aims overall related technology development and obtainment of related license in 2001. The development direction is to determine the reactor type and to build up the design concept in 1994. For development trend analysis of foreign next generation reactor, level-1 PSA, fuel cycle analysis and computer code development are performed on System 80+ and AP 600. Especially for design characteristics analysis and volume upgrade of AP 600, nuclear fuel and reactor core design analysis, coolant circuit design analysis, mechanical structure design analysis and safety analysis etc. are performed. (Author).

İbn Kuteybe’ye Göre Şiirde Yetenek Sorunu

OpenAIRE

Öznurhan, Halim

2005-01-01

‛Abdü’s-Selâm ‛Abdü’l-Hafîz ‛Abdü’l-‛Âl, Nakdü’ş-şi‛r beyne İbn Kuteybe ve İbn Tabâtabâ el-‛Alevî, Dâru’l-fikri’l-‛Arabî, Mısır, ts. , 166. ‛Abdü’l-‛Âl, aynı yer. İbn Kuteybe, 12, 13. Ebû ‛Usmân ‛Amr b. Bahr el-Câhiz, el-Beyân ve’t-tebyîn, nşr: ‛Abdü’s-Selâm M. Hârûn, Kahire, 1948, I, 208. İbn Kuteybe, 17. Krş: el-Câhiz, I, 207 ve II, 13. Krş: el-Câhiz, II, 13. Bkz: İbn Kuteybe, 476, 497, 501

Integrated Guidelines for Management of Alloy 600 Locations

Energy Technology Data Exchange (ETDEWEB)

Na, Kyung-Hwan; Chung, Hansub; Yang, Jun-Seog; Lee, Kyoung-Soo [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The locations experiencing PWSCC include steam generator tubes, pressurizer instrumental nozzles, control rod driving mechanism(CRDM) penetration nozzles, reactor outlet nozzles, and bottom mounted instrumental(BMI) nozzles. Korea Hydro and Nuclear Power Co.(KHNP) has developed integrated guidelines for management of alloy 600 locations and the guidelines are under review by the regulator. The guidelines consist of alloy 600 location database, inspection program, maintenance/preventive maintenance method, and finally water chemistry management for PWSCC mitigation. In this paper, the detailed contents are presented. The integrated guidelines collected all relevant information on the management of alloy 600 locations. This information may be useful for establishing the most effective preventive maintenance strategies by prioritization in addition to maintenance strategies. Table II summarize maintenance strategies for alloy 600 locations.

Advanced light-water reactors

International Nuclear Information System (INIS)

Golay, M.W.; Todreas, N.E.

1990-01-01

Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Catana, Alexandru [RAAN, Institute for Nuclear Research, Str. Campului Nr. 1, Pitesti, Arges (Romania); Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel [University POLITEHNICA of Bucharest (Romania)

2008-07-01

Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D{sub 2}O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D{sub 2}O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

International Nuclear Information System (INIS)

Catana, Alexandru; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

2008-01-01

Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D 2 O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D 2 O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

[Radiation ecological environment in the Republic of Kazakhstan in the vicinity of the reactors and on the territory of the Semipalatinsk Test Site].

Science.gov (United States)

Kim, D S

2012-01-01

The results of research into the environmental conditions in the regions of location of the pressurized water reactor WWR-K, fast neutron breeder BN-350 and on the territory of the Semipalatinsk Test Site are represented. The effects of the exposure to aerosol emissions from WWR-K and BN-350 reactors on the environment are summarized. We present some arguments in favor of the safe operation of fission reactors in compliance with the rules and norms of nuclear and radiation protection and the efficient disposal of radioactive waste on the territory of the Republic.

Candu 600 fuelling machine testing, the romanian experience

International Nuclear Information System (INIS)

Valeca, S.; Doca, C.; Iorga, C.

2013-01-01

The Candu 600 Fuelling Machine is a complex mechanism which must run in safety conditions and with high reliability in the Candu Reactor. The testing and commissioning process of this nuclear equipment meets the high standards of NPPs requirements using special technological facilities, modern measurement instruments as well the appropriate IT resources for data acquisition and processing. The paper presents the experience of the Institute for Nuclear Research Pitesti, Romania, in testing Candu 600 Fuelling Machines, inclusive the implied facilities, and in development of four simulators: two dedicated for the training of the Candu 600 Fuelling Machine Operators, and another two to simulate some process signals and actions. (authors)

Basic research on mechanism of BN inclusion in improving the machinability of steel

International Nuclear Information System (INIS)

Ya-nan, C.; Yan-ping, B.; Min, W.; Xiao-feng, C.; Lin-jing, W.; Li-hua, Z.

2014-01-01

Boron nitride-added eco-friendly free cutting steel has recently drawn more and more attention. But, the mechanisms explaining the role of BN inclusions improving the machinability of steels is not very clear. In this investigation, the material removal mechanism for cutting of BN inclusions in steels is explored, using a combination of theoretical analysis and a series of experiments. First, the actual shape of BN inclusions is observed and the amount and distribution of BN inclusions is quantitatively analyzed. Subsequently, the cutting performance of the steel is determined by cutting experimental tests. Moreover, the micro mechanical properties and the material removal mechanisms for cutting of BN inclusions are investigated by means of nano indentation. The results revealed that the BN inclusions are hexagonal and are uniformly distributed, their average content is 23.2 per unit area and their volume fraction is 0.51% in the steel with 74 ppm B and 180 ppm N. It is shown that BN inclusions can improve the cutting performance of steel significantly, and a model describing the material removal mechanism for cutting of BN inclusions is proposed. BN inclusions act as stress concentration source, lubrication and wrap page of hard particles. (Author)

Basic research on mechanism of BN inclusion in improving the machinability of steel

Energy Technology Data Exchange (ETDEWEB)

Ya-nan, C.; Yan-ping, B.; Min, W.; Xiao-feng, C.; Lin-jing, W.; Li-hua, Z.

2014-07-01

Boron nitride-added eco-friendly free cutting steel has recently drawn more and more attention. But, the mechanisms explaining the role of BN inclusions improving the machinability of steels is not very clear. In this investigation, the material removal mechanism for cutting of BN inclusions in steels is explored, using a combination of theoretical analysis and a series of experiments. First, the actual shape of BN inclusions is observed and the amount and distribution of BN inclusions is quantitatively analyzed. Subsequently, the cutting performance of the steel is determined by cutting experimental tests. Moreover, the micro mechanical properties and the material removal mechanisms for cutting of BN inclusions are investigated by means of nano indentation. The results revealed that the BN inclusions are hexagonal and are uniformly distributed, their average content is 23.2 per unit area and their volume fraction is 0.51% in the steel with 74 ppm B and 180 ppm N. It is shown that BN inclusions can improve the cutting performance of steel significantly, and a model describing the material removal mechanism for cutting of BN inclusions is proposed. BN inclusions act as stress concentration source, lubrication and wrap page of hard particles. (Author)

Scanning tunneling microscopy of hexagonal BN grown on graphite

International Nuclear Information System (INIS)

Fukumoto, H.; Hamada, T.; Endo, T.; Osaka, Y.

1991-01-01

The microscopic surface topography of thin BN x films grown on graphite by electron cyclotron resonance plasma chemical vapor deposition have been imaged with scanning tunneling microscopy in air. The scanning tunneling microscope has generated images of hexagonal BN with atomic resolution

Irradiation temperature dependence of defect formation of nitrides (A1N and c-BN) during neutron irradiations

International Nuclear Information System (INIS)

Atobe, Kozo.; Okada, Moritami; Nakagawa, Masuo

2000-01-01

The nitrogen vacancy concentration in the more refractory nitrides (A1N and c-BN) is determined as a function of reactor fluence up to 5.2x10 17 thermal neutrons/cm 2 and a function of the irradiation temperature at 25, 50, 100, 150, 200, 250 K. It is found that there is no remarkable dependence of the defect formation in nitrides on the irradiation temperature. The production of damage in the nitrides is considerably different from that in oxides. From the irradiation experiments using thermal neutron irradiation field, it is suggested in reactor irradiation that the atomic displacements in the nitrides occur predominately from energetic particles of the nuclear reactions with thermal neutrons in addition to the elastic collisions by fast neutron

Introducing lattice strain to graphene encapsulated in hBN

Science.gov (United States)

Tomori, Hikari; Hiraide, Rineka; Ootuka, Youiti; Watanabe, Kenji; Taniguchi, Takashi; Kanda, Akinobu

Due to the characteristic lattice structure, lattice strain in graphene produces an effective gauge field. Theories tell that by controlling spatial variation of lattice strain, one can tailor the electronic state and transport properties of graphene. For example, under uniaxial local strain, graphene exhibits a transport gap at low energies, which is attractive for a graphene application to field effect devices. Here, we develop a method for encapsulating a strained graphene film in hexagonal boron-nitride (hBN). It is known that the graphene carrier mobility is significantly improved by the encapsulation of graphene in hBN, which has never been applied to strained graphene. We encapsulate graphene in hBN using the van der Waals assembly method. Strain is induced by sandwiching a graphene film between patterned hBN sheets. Spatial variation of strain is confirmed with micro Raman spectroscopy. Transport measurement of encapsulated strained graphene is in progress.

I-V characteristics of graphene nanoribbon/h-BN heterojunctions and resonant tunneling.

Science.gov (United States)

Wakai, Taiga; Sakamoto, Shoichi; Tomiya, Mitsuyoshi

2018-07-04

We present the first principle calculations of the electrical properties of graphene sheet/h-BN heterojunction (GS/h-BN) and 11-armchair graphene nanoribbon/h-BN heterojunction (11-AGNR/h-BN), which are carried out using the density functional theory (DFT) method and the non-equilibrium Green's function (NEGF) technique. Since 11-AGNR belongs to the conductive (3n-1)-family of AGNR, both are metallic nanomaterials with two transverse arrays of h-BN, which is a wide-gap semi-conductor. The two h-BN arrays act as double barriers. The transmission functions (TF) and I-[Formula: see text] characteristics of GS/h-BN and 11-AGNR/h-BN are calculated by DFT and NEGF, and they show that quantum double barrier tunneling occurs. The TF becomes very spiky in both materials, and it leads to step-wise I-[Formula: see text] characteristics rather than negative resistance, which is the typical behavior of double barriers in semiconductors. The results of our first principle calculations are also compared with 1D Dirac equation model for the double barrier system. The model explains most of the peaks of the transmission functions nearby the Fermi energy quite well. They are due to quantum tunneling.

Current status of development in dry pyro-electrochemical technology of SNF reprocessing

International Nuclear Information System (INIS)

Bychkov, A.V.; Skiba, O.V.; Kormilitsyn, M.V.

2004-01-01

The technology of SNF management in molten salts currently developed by a group of institutes headed by RIAR has had several stages of development: - basic research of uranium, plutonium and main FP properties (investigation and reprocessing of different kinds of SNF in 1960 - 1970); - development of the equipment and implementation of the pyro-electrochemical technology of granulated UPu fuel production. Development of the vibro-packing method and in-pile testing of vibro-packed fuel pins with granulated fuel as the most 'logical' continuation of reprocessing: implementation of the technology for BOR-60 and BN-600 (1980 - 1990); - development of closed fuel cycle elements. Checking of the technology using batches of SNF. In-pile tests. Feasibility study of the closed fuel cycle (CFC). Study of application of the technology to other objects (transmutation; nitride, cermet and other fuels) (1980 - 1990). The current status of the research is the following: - Basic research. Properties of uranium, plutonium, thorium, and neptunium in chloride melts have been studied in much detail. The data on physical chemistry and electrochemistry of the main FP is enough for understanding the processes. Detailed studies of americium, curium, and technetium chemistry are the essential investigation directions; - Engineering development. The technology and equipment bases have been developed for the processes of oxide fuel reprocessing and fabrication. The technology was checked using 5500 kg of pure fuel from different reactors and 20 kg of irradiated BN-350 and BOR-60 fuel. The bases of the technology have been provided and the feasibility study has been carried out for a full-scale plant of BN-800 CFC; - Industrial application: Since the technology is highly prepared, the activities on industrial application of U-Pu fuel are now underway. The BOR-60 reactor uses fuel obtained by the dry method, the design of the facility for implementation of CFC reactors is being developed. 9

Objectives and status of development of AC600

International Nuclear Information System (INIS)

Zhao Chengkun

1997-01-01

AC600 is a medium power capability nuclear power station of next generation, which is developed based on world nuclear power improving tendency, requirements of custom with considering China situation and technical foundation. Its main technical characteristics are as following: advanced core and passive safety system, double loop standard design and international popular equipment. Meanwhile, it a simplification of present system, using advanced control room and pattern construction thus developed the operation reliability of nuclear power station, lower construction and operating cost. In order to accelerate the development of next generation advanced reactor, cooperating with Westinghouse Electric Corporation, the joint economic technical research has been established. Based on AC600, the CAP600 is developed on further improving safety and reliability, economical and electric network adoption of AC600

Orthorhombic BN: A novel superhard sp{sup 3} boron nitride allotrope

Energy Technology Data Exchange (ETDEWEB)

Zhang, Zhiguo [College of Physics, Beihua University, Jilin 132013 (China); Lu, Mingchun [Department of Aeronautical Engineering Professional Technology, Jilin Institute of Chemical Technology, Jilin 132102 (China); Zhu, Li; Zhu, Lili; Li, Yadan [College of Physics, Beihua University, Jilin 132013 (China); Zhang, Miao, E-mail: zhangmiaolmc@126.com [College of Physics, Beihua University, Jilin 132013 (China); College of Materials Science and Engineering, National Laboratory of Superhard Materials, Jilin University, Changchun 130012 (China); Li, Quan, E-mail: liquan777@jlu.edu.cn [College of Materials Science and Engineering, National Laboratory of Superhard Materials, Jilin University, Changchun 130012 (China)

2014-02-07

Here, a novel superhard orthorhombic allotrope of boron nitride (O-BN) with the space group of Pbam has been predicted using first-principles calculations. Our results revealed that O-BN simultaneously posses incompressible with a high bulk modulus of 397.38 GPa, and superhard properties with a high Vickers hardness of 65 GPa. Further phonon calculations show O-BN structure is dynamically stable. Moreover, it is thermodynamics energetically more preferable than previous proposed BN allotropes and a transparent insulator with an indirect band gap of about 4.85 eV. Our researches represent a significant step toward the exploration of superhard materials.

Experience and results of material science research conducted on spent fuel assemblies from the BN-350 fast reactor

Energy Technology Data Exchange (ETDEWEB)

Maksimkin, O.; Gusev, M.; Turubarova, L.G.; Tsai, K.V.; Yarovchuk, A.V. [Institute of Nuclear Physics, Almaty (Kazakhstan)

2007-07-01

Full text of publication follows: The BN-350 fast reactor was commissioned in 1973, ran successfully for many years and is now in the decommission stage. Its unique operational parameters (low temperature of sodium at the input, wide range of damage rates, etc. ) allowed the investigation of a number of new radiation effects on both austenitic and ferritic-martensitic steels. The latter class of steel was extensively employed as wrappers for fuel assemblies. Much of the accumulated experience in BN-350 is relevant to development of fusion devices. Results are presented on post-operational research of steels 12Cr18Ni10Ti, 08Cr16Ni11Mo3, and 12Cr13Mo2BFR, all serving as hexagonal shrouds of fuel assemblies. Structural materials in the active core zone operated at temperatures of 280-430 deg. C, and were irradiated the range of 0.25-83 dpa with damage rates of 10{sup -9} - 10{sup -6} dpa/s). Investigations of irradiated hexagonal shroud materials were performed with using traditional techniques of transmission and scanning electron microscopy, metallography, mechanical tests, hydrostatic weighing, magnetometry, etc. Additionally, new techniques have been developed and employed with great success on these highly irradiated materials, such as optical computer extensometry, and magnetization cartography. Typical results to be covered in this presentation are: a) In 12Cr18Ni10Ti steel irradiated at a low dose rate of 0.12 x 10{sup -8} dpa/s voids were found at 281 deg. C after only 0.65 dpa, demonstrating once again the acceleration of swelling at low dpa rates observed in other steels. b) Data on helium release during annealing of highly irradiated sample are presented. c) Differences in deformation-induced hardening between the shroud's corners and faces leads to post-irradiation differences in swelling and mechanical properties. d) During room temperature mechanical tests of 12Cr18Ni10Ti steel at {approx}56 dpa at 350 deg. C it was found that ductility lost at

AP600 containment purge radiological analysis

Energy Technology Data Exchange (ETDEWEB)

O`Connor, M.; Schulz, J.; Tan, C. [Bechtel Power Corporation (United States)] [and others

1995-02-01

The AP600 Project is a passive pressurized water reactor power plant which is part of the Design Certification and First-of-a-Kind Engineering effort under the Advanced Light Water Reactor program. Included in this process is the design of the containment air filtration system which will be the subject of this paper. We will compare the practice used by previous plants with the AP600 approach to meet the goals of industry standards in sizing the containment air filtration system. The radiological aspects of design are of primary significance and will be the focus of this paper. The AP600 Project optimized the design to combine the functions of the high volumetric flow rate, low volumetric flow rate, and containment cleanup and other filtration systems into one multi-functional system. This achieves a more simplified, standardized, and lower cost design. Studies were performed to determine the possible concentrations of radioactive material in the containment atmosphere and the effectiveness of the purge system to keep concentrations within 10CFR20 limits and within offsite dose objectives. The concentrations were determined for various reactor coolant system leakage rates and containment purge modes of operation. The resultant concentrations were used to determine the containment accessibility during various stages of normal plant operation including refueling. The results of the parametric studies indicate that a dual train purge system with a capacity of 4,000 cfm per train is more than adequate to control the airborne radioactivity levels inside containment during normal plant operation and refueling, and satisfies the goals of ANSI/ANS-56.6-1986 and limits the amount of radioactive material released to the environment per ANSI/ANS 59.2-1985 to provide a safe environment for plant personnel and offsite residents.

BnNHL18A shows a localization change by stress-inducing chemical treatments

International Nuclear Information System (INIS)

Lee, Suk-Bae; Ham, Byung-Kook; Park, Jeong Mee; Kim, Young Jin; Paek, Kyung-Hee

2006-01-01

The two genes, named BnNHL18A and BnNHL18B, showing sequence homology with Arabidopsis NDR1/HIN1-like (NHL) genes, were isolated from cDNA library prepared with oilseed rape (Brassica napus) seedlings treated with NaCl. The transcript level of BnNHL18A was increased by sodium chloride, ethephon, hydrogen peroxide, methyl jasmonate, or salicylic acid treatment. The coding regions of BnNHL18A and BnNHL18B contain a sarcolipin (SLN)-like sequence. Analysis of the localization of smGFP fusion proteins showed that BnNHL18A is mainly localized to endoplasmic reticulum (ER). This result suggests that the SLN-like sequence plays a role in retaining proteins in ER membrane in plants. In response to NaCl, hydrogen peroxide, ethephon, and salicylic acid treatments, the protein localization of BnNHL18A was changed. Our findings suggest a common function of BnNHL18A in biotic and abiotic stresses, and demonstrate the presence of the shared mechanism of protein translocalization between the responses to plant pathogen and to osmotic stress

Optoelectronic properties of higher acenes, their BN analogue and substituted derivatives

International Nuclear Information System (INIS)

Armaković, Stevan; Armaković, Sanja J.; Holodkov, Vladimir; Pelemiš, Svetlana

2016-01-01

We have investigated optoelectronic properties of higher acenes: pentacene, hexacene, heptacene, octacene, nonacene, decacene and their boron-nitride (BN) analogues, within the framework of density functional theory (DFT). We have also investigated the optoelectronic properties of acenes modified by BN substitution. Calculated optoelectronic properties encompasses: oxidation and reduction potentials, electron and hole reorganization energies and energy difference between excited first singlet and triplet states ΔE(S_1−T_1). Oxidation and reduction potentials indicate significantly better stability of BN analogues, comparing with their all-carbon relatives. Although higher acenes possess lower electron and hole reorganization energies, with both best values much lower than 0.1 eV, their BN analogues also have competitive values of reorganization energies, especially for holes for which reorganization energy is also lower than 0.1 eV. On the other hand ΔE(S_1−T_1) is much better for BN analogues, having values that indicate that BN analogues are possible applicable for thermally activated delayed fluorescence. - Highlights: • Optoelectronic properties of structures based on higher acenes have been investigated. • Oxidation and reduction potentials together with reorganization energies are calculated. • TADF is analyzed through calculation of ΔE(S_1−T_1), which is much better for BN analogues. • Reorganization energies of acenes improve with the increase of number of benzene rings.

A comparative computational study on the BN ring doped nanographenes

Energy Technology Data Exchange (ETDEWEB)

Vessally, E. [Department of Chemistry, Payame Noor University, Tehran (Iran, Islamic Republic of); Soleimani-Amiri, S. [Department of Chemistry, Karaj Branch, Islamic Azad University, Karaj (Iran, Islamic Republic of); Hosseinian, A. [Department of Engineering Science, College of Engineering, University of Tehran, P.O. Box 11365-4563, Tehran (Iran, Islamic Republic of); Edjlali, L., E-mail: l_edjlali@iaut.ac.ir [Department of Chemistry, Tabriz Branch, Islamic Azad University, Tabriz (Iran, Islamic Republic of); Bekhradnia, A. [Pharmaceutical Sciences Research Center, Department of Medicinal Chemistry, Mazandaran University of Medical Sciences, Sari (Iran, Islamic Republic of)

2017-02-28

Highlights: • Clar’s sextet rule determine the relative stability of HBC nanographenes. • Coronene-like doping increases the electrical conductivity of the HBC. • Frenkel type exciton binding energy is predicted for HBC nanographenes. - Abstract: The electronic, optical, energetic, and structural properties of a HBC (hexa-peri-hexabenzocoronene) nanographene and its central benzene- and coronene-like BN substituted forms, and also full BN analogue were investigated using density functional theory. It was found that a larger number of carbon atoms cause a more negative cohesive energy and, thereby a greater structural stability. Our nucleus independent chemical shift analysis indicates that the aromaticity and Clar’s sextet rule determine the relative stability of these structures. The benzene-like or coronene-like doping makes the HBC more insulator or semiconductor. Electron-hole Frenkel type exciton binding energy was predicted and calculated to be nearly identical for all nanographenes in the range of 0.61–0.69 eV. The coronene-like BN-doped HBC (BN2-HBN) shows higher conductivity due to very narrow optical and HOMO-LUMO energy gap. Partial density of states analysis indicates that the BN2-HBC electronically can be assumed a full BN whose peripheral atoms are replaced by carbon atoms. These carbon atoms are responsible for new states which are appeared within the gap.

Electronic properties of T graphene-like C-BN sheets: A density functional theory study

Science.gov (United States)

Majidi, R.

2015-11-01

We have used density functional theory to study the electronic properties of T graphene-like C, C-BN and BN sheets. The planar T graphene with metallic property has been considered. The results show that the presence of BN has a considerable effect on the electronic properties of T graphene. The T graphene-like C-BN and BN sheets show semiconducting properties. The energy band gap is increased by enhancing the number of BN units. The possibility of opening and controlling band gap opens the door for T graphene in switchable electronic devices.

Basic research on mechanism of BN inclusion in improving the machinability of steel

Directory of Open Access Journals (Sweden)

Ya-nan, Chen

2014-12-01

Full Text Available Boron nitride-added eco-friendly free cutting steel has recently drawn more and more attention. But, the mechanisms explaining the role of BN inclusions improving the machinability of steels is not very clear. In this investigation, the material removal mechanism for cutting of BN inclusions in steels is explored, using a combination of theoretical analysis and a series of experiments. First, the actual shape of BN inclusions is observed and the amount and distribution of BN inclusions is quantitatively analyzed. Subsequently, the cutting performance of the steel is determined by cutting experimental tests. Moreover, the micro mechanical properties and the material removal mechanisms for cutting of BN inclusions are investigated by means of nanoindentation. The results revealed that the BN inclusions are hexagonal and are uniformly distributed, their average content is 23.2 per unit area and their volume fraction is 0.51% in the steel with 74 ppm B and 180 ppm N. It is shown that BN inclusions can improve the cutting performance of steel significantly, and a model describing the material removal mechanism for cutting of BN inclusions is proposed. BN inclusions act as stress concentration source, lubrication and wrappage of hard particles.Los aceros de fácil mecanizado o corte libre con nitruro de boro agregado han despertado un gran interés. Sin embargo, aún no se han determinado los mecanismos que explican el papel de las inclusiones de BN en la mejora de la maquinabilidad de estos aceros. En este trabajo, se investigan los mecanismos de corte de las inclusiones BN en aceros mediante la combinación de un análisis teórico y una serie de experimentos. En primer lugar, se determina la morfología de las inclusiones BN y se analiza cuantitativamente la cantidad y distribución de las mismas. Posteriormente, el rendimiento de corte del acero se determina mediante ensayos de corte. Por otra parte, las propiedades mecánicas locales y los

Microstructure and mechanical properties of SiO2-BN ceramic and Invar alloy joints brazed with Ag–Cu–Ti+TiH2+BN composite filler

Directory of Open Access Journals (Sweden)

Y. Wang

2016-03-01

Full Text Available Ag–Cu–Ti + TiH2+BN composite filler was prepared to braze SiO2-BN ceramic and Invar alloy. The interfacial microstructure, mechanical properties, and residual stress distribution of the brazed joints were investigated. The results show that a wave-like Fe2Ti–Ni3Ti structure appears in the Invar substrate and a thin TiN–TiB2 reaction layer forms adjacent to the SiO2-BN ceramic. The added BN particles react with Ti to form TiN–TiB fine-particles, which is beneficial to refine the microstructure of the brazing seam and to greatly inhibit the brittle compounds formation. The interfacial microstructure at various brazing temperatures was analyzed, and the mechanism for the interfacial reactions responsible for the bonding was proposed. The maximum shear strength of the joints brazed with the composite filler at 880 °C for 10 min is 39 MPa, which is 30% greater than that brazed with Ag–Cu–Ti alloy. The improvement of the joint strength is attributed to the variation of joint microstructure and the reduction of tensile stresses induced in the SiO2-BN ceramic. The finite element analysis indicates that the peak tensile stress decreases from 230 to 142 MPa due to the addition of BN particles in the ceramic.

Loss-of-normal-feedwater sensitivity studies for AP600 behavior characterization

International Nuclear Information System (INIS)

Saiu, G.

1996-01-01

Activity concerning the development of a RELAP5/MOD3 model to simulate the Westinghouse Electric Corporation AP600 is summarized. The aim is to gain initial insight into the capability of RELAP5 to simulate the behavior of AP600 safety features. A-loss-of-normal-feedwater event is studied. Of the transients that must be investigated, this transient has been chosen to be one of the most relevant because the response of the AP600 to a loss-of-normal-feedwater event differs significantly from that of current pressurized water reactors in the extensive use of passive safety features peculiar to the AP600. Also, strong interactions among the AP600 safety systems, which should be further analyzed to permit full optimization of the system actuation logic and operation, are shown. Finally, a loss of normal feedwater without reactor scram, performed to investigate short-term plant behavior, shows that the pressure peak is affected by critical discharge flow coefficients applied to the pressurizer safety valves, while a relatively small reduction of the pressure peak is observed when both heat exchangers of the passive heat removal system are operating as opposed to the case in which only one is available. The data used for this study are derived from the Standard Safety Analysis Report configuration of the Westinghouse AP600 as of 1992

Irradiation creep and stress-enhanced swelling of Fe-16Cr-15Ni-Nb austenitic stainless steel in BN-350

Energy Technology Data Exchange (ETDEWEB)

Vorobjev, A.N.; Porollo, S.I.; Konobeev, Yu.V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)] [and others

1997-04-01

Irradiation creep and void swelling will be important damage processes for stainless steels when subjected to fusion neutron irradiation at elevated temperatures. The absence of an irradiation device with fusion-relevant neutron spectra requires that data on these processes be collected in surrogate devices such as fast reactors. This paper presents the response of an annealed austenitic steel when exposed to 60 dpa at 480{degrees}C and to 20 dpa at 520{degrees}C. This material was irradiated as thin-walled argon-pressurized tubes in the BN-350 reactor located in Kazakhstan. These tubes were irradiated at hoop stresses ranging from 0 to 200 MPa. After irradiation both destructive and non-destructive examination was conducted.

Predictive methodology to address PWSCC of Alloy 600 locations in PWRS

International Nuclear Information System (INIS)

Rao, G.V.

1992-01-01

Contributing factors to primary water stress corrosion cracking (PWSCC) are susceptible microstructure, temperature, and residual and applied stresses. In order to predict PWSCC of Inconel 600 components in PWR type reactors, a number of steps were taken. All Inconel 600 components were located, fabrication history, weld procedures and material properties were identified. Service temperatures and approximate stresses were determined. Precise service stress evaluations of Inconel 600 locations by Finite Element and other analytical evaluations were made. Using data analysis, relative PWSCC susceptibility evaluations of Inconel 600 locations were made on the basis of the Westinghouse RSI model. Finally, a prioritized inspection plan for Inconel 600 locations was developed and recommendations provided. 11 figs., 2 tabs

Optoelectronic properties of higher acenes, their BN analogue and substituted derivatives

Energy Technology Data Exchange (ETDEWEB)

Armaković, Stevan, E-mail: stevan.armakovic@df.uns.ac.rs [University of Novi Sad, Faculty of Sciences, Department of Physics, Trg Dositeja Obradovića 4, 21000, Novi Sad (Serbia); Armaković, Sanja J. [University of Novi Sad, Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg Dositeja Obradovića 3, 21000, Novi Sad (Serbia); Holodkov, Vladimir [Educons University, Faculty of Sport and Tourism - TIMS, Radnička 30a, 21000, Novi Sad (Serbia); Pelemiš, Svetlana [University of East Sarajevo, Faculty of Technology, Karakaj bb, 75400, Zvornik, Republic of Srpska, Bosnia and Herzegovina (Bosnia and Herzegovina)

2016-02-15

We have investigated optoelectronic properties of higher acenes: pentacene, hexacene, heptacene, octacene, nonacene, decacene and their boron-nitride (BN) analogues, within the framework of density functional theory (DFT). We have also investigated the optoelectronic properties of acenes modified by BN substitution. Calculated optoelectronic properties encompasses: oxidation and reduction potentials, electron and hole reorganization energies and energy difference between excited first singlet and triplet states ΔE(S{sub 1}−T{sub 1}). Oxidation and reduction potentials indicate significantly better stability of BN analogues, comparing with their all-carbon relatives. Although higher acenes possess lower electron and hole reorganization energies, with both best values much lower than 0.1 eV, their BN analogues also have competitive values of reorganization energies, especially for holes for which reorganization energy is also lower than 0.1 eV. On the other hand ΔE(S{sub 1}−T{sub 1}) is much better for BN analogues, having values that indicate that BN analogues are possible applicable for thermally activated delayed fluorescence. - Highlights: • Optoelectronic properties of structures based on higher acenes have been investigated. • Oxidation and reduction potentials together with reorganization energies are calculated. • TADF is analyzed through calculation of ΔE(S{sub 1}−T{sub 1}), which is much better for BN analogues. • Reorganization energies of acenes improve with the increase of number of benzene rings.

The Westinghouse AP600 an advanced nuclear option for small or medium electricity grids

International Nuclear Information System (INIS)

Bruschi, H. J.; Novak, V.

1996-01-01

During the early days of commercial nuclear power, many countries looking to add nuclear power to their energy mix required large plants to meet the energy needs of rapidly growing populations and large industrial complexes. The majority of plants worldwide are in the range of 100 megawatts and beyond. During the 1970s, it became apparent that a smaller nuclear plants would appeal to utilities looking to add additional power capacity to existing grids, or to utilities in smaller countries which were seeking efficient, new nuclear generation capacity for the first time. For instance, the Westinghouse-designed 600 megawatt Krsko plant in Slovenia began operation in 1980, providing electricity to inhabitants of relatively small, yet industrial populations of Slovenia and Croatia. This plant design incorporated the best, proven technology available at that time, based on 20 years of Westinghouse PWR pioneering experience. Beginning in the early 1980s, Westinghouse began to build further upon that experience - in part through the advanced light water reactor programs established by the Electric Power Research institute (EPRI) and the U.S. Department of Energy (DOE) - to design a simplified, advanced nuclear reactor in the 600 megawatt range. Originally, Westinghouse's development of its AP600 (advanced, passive 600-megawatt) plants was geared towards the needs of U.S. utilities which specified smaller, simplified nuclear options for the decades ahead. It soon became evident that the small and medium sized electricity grids of international markets could benefit from this new reactor. From the earliest days of Westinghouse's AP600 development, the corporation invited members of the international nuclear community to take part in the design, development and testing of the AP600 - with the goal of designing a reactor that would meet the diverse needs of an international industry composed of countries with similar, yet different, concerns. (author)

Phase stability limit of c-BN under hydrostatic and non-hydrostatic pressure conditions

International Nuclear Information System (INIS)

Xiao, Jianwei; Du, Jinglian; Wen, Bin; Zhang, Xiangyi; Melnik, Roderick; Kawazoe, Yoshiyuki

2014-01-01

Phase stability limit of cubic boron nitride (c-BN) has been investigated by the crystal structure search technique. It indicated that this limit is ∼1000 GPa at hydrostatic pressure condition. Above this pressure, c-BN turns into a metastable phase with respect to rocksalt type boron nitride (rs-BN). However, rs-BN cannot be retained at 0 GPa owing to its instability at pressure below 250 GPa. For non-hydrostatic pressure conditions, the phase stability limit of c-BN is substantially lower than that under hydrostatic pressure conditions and it is also dramatically different for other pressure mode

Structural analysis of graphene and h-BN: A molecular dynamics approach

International Nuclear Information System (INIS)

Thomas, Siby; Ajith, K. M.; Valsakumar, M. C.

2016-01-01

Classical molecular dynamics simulation is employed to analyze pair correlations in graphene and h-BN at various temperatures to explore the integrity of their respective structures. As the temperature increases, the height fluctuations in the out-of-plane direction of both graphene and h-BN are found to increase. The positional spread of atoms also increases with temperature. Thus the amplitude of the peak positions in the radial distribution function (RDF) decreases with temperature. It is found that FWHM of peaks in the RDF of h-BN is smaller as compared to those of graphene which implies that the structure of h-BN is more robust as compared to that of graphene with respect to their respective empirical potential.

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)

Grumer, U.

1990-06-01

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

Electronic structure properties of deep defects in hBN

Science.gov (United States)

Dev, Pratibha; Prdm Collaboration

In recent years, the search for room-temperature solid-state qubit (quantum bit) candidates has revived interest in the study of deep-defect centers in semiconductors. The charged NV-center in diamond is the best known amongst these defects. However, as a host material, diamond poses several challenges and so, increasingly, there is an interest in exploring deep defects in alternative semiconductors such as hBN. The layered structure of hBN makes it a scalable platform for quantum applications, as there is a greater potential for controlling the location of the deep defect in the 2D-matrix through careful experiments. Using density functional theory-based methods, we have studied the electronic and structural properties of several deep defects in hBN. Native defects within hBN layers are shown to have high spin ground states that should survive even at room temperature, making them interesting solid-state qubit candidates in a 2D matrix. Partnership for Reduced Dimensional Material (PRDM) is part of the NSF sponsored Partnerships for Research and Education in Materials (PREM).

Li{sub 4}Ba[BN{sub 2}]{sub 2} - structure and vibrational spectra

Energy Technology Data Exchange (ETDEWEB)

Seidel, Stefan; Rodewald, Ute C.; Poettgen, Rainer [Institut fuer Anorganische und Analytische Chemie, Universitaet Muenster (Germany); Somer, Mehmet; Kiraz, Kamil [Chemistry Department, Koc University, Sariyer-Istanbul (Turkey)

2017-12-13

The nitridoborate Li{sub 4}Ba[BN{sub 2}]{sub 2} was synthesized from a 4:1 molar ratio of Li{sub 3}[BN]{sub 2} and Ba{sub 3}[BN{sub 2}]{sub 2} in an arc-welded niobium ampoule at a maximum annealing temperature of 1173 K. The structure was refined from single-crystal X-ray diffractometer data: new type, P1, a = 533.9(2), b = 585.0(3), c = 860.6(4) pm, α = 80.72(3), β = 73.84(6), γ = 89.87(4) , wR{sub 2} = 0.1196, 1429 F{sup 2} values, 50 variables. The Li{sub 4}Ba[BN{sub 2}]{sub 2} structure contains two crystallographically independent [BN{sub 2}]{sup 3-} units with 134 pm B-N distance, which are slightly bent: 178 for N2-B1-N1 and 175 for N4-B2-N3. Due to the high lithium content both [BN{sub 2}]{sup 3-} units have a strongly distorted coordination by 8Li{sup +} + 3Ba{sup 2+}. The four crystallographically independent lithium cations show distorted tetrahedral coordination by [BN{sub 2}]{sup 3-} units with Li-N distances ranging from 195 to 247 pm. IR and Raman spectra show the typical vibrations of the [BN{sub 2}] unit along with a well-resolved splitting of the ν({sup 10}B) and ν({sup 11}B) frequencies. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

AP600 large-break loss-of-collant-accident developmental assessment plan for TRAC-PF1/MOD2

International Nuclear Information System (INIS)

Knight, T.D.

1996-07-01

The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA

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.

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

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2006-06-01

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

Steps to Advanced CANDU 600

International Nuclear Information System (INIS)

Oh, Yongshick; Brooks, G. L.

1988-01-01

The CANDU nuclear power system was developed from merging of AECL heavy water reactor technology with Ontario Hydro electrical power station expertise. The original four units of Ontario Hydro's Pickering Generating Station are the first full-scale commercial application of the CANDU system. AECL and Ontario Hydro then moved to the next evolutionary step, a more advanced larger scale design for four units at the Bruce Generating Station. CANDU 600 followed as a single unit nuclear electric power station design derived from an amalgam of features of the multiple unit Pickering and Bruce designs. The design of the CANDU 600 nuclear steam supply system is based on the Pickering design with improvements derived from the Bruce design. For example, most CANDU 600 auxiliary systems are based on Bruce systems, whereas the fuel handling system is based on the Pickering system. Four CANDU 600 units are in operation, and five are under construction in Romania. For the additional four units at Pickering Generating Station 'B', Ontario Hydro selected a replica of the Pickering 'A' design with limited design changes to maintain a high level of standardization across all eight units. Ontario Hydro applied a similar policy for the additional four units at Bruce Generating Station 'B'. For the four unit Darlington station, Ontario Hydro selected a design based on Bruce with improvements derived from operating experience, the CANDU 600 design and development programs

PG BN 1600 sodium fire protection system

International Nuclear Information System (INIS)

Bar, J.; Urbancik, L.

1978-12-01

A design was developed of a fire protection system for steam generator of a 1600 MW sodium cooled fast reactor (BN-1600). Chemical reactions are described of liquid sodium with atmospheric components and solid materials coming into contact with sodium in its release from the steam generator, and in safeguarding protection against sodium fires. The requirements for the purity of nitrogen as an atmosphere inert to liquid sodium are given. Characteristics and basic parameters are shown of level and spray fires, elementary terms are explained concerning the properties of aerosols formed during fires, the methods and means of release signalling and fire alarm are described as are fire precautions using fire-fighting equipment, modifying the support tank and the cell bottom and building sewage pits. The design of the system comprises an alarm system for liquid sodium using point and line electric contact sensors and flame photometer based aerosol sensors as well as a fire-fighting system based on the system of channelling liquid sodium into emergency discharge tanks filled with an inert gas, a set of fire extinguishers and other fire fighting material, and measures for the elimination of sodium fire consequences. (J.B.)

Preliminary Economic Assessment of KALIMER-600

International Nuclear Information System (INIS)

Moon, Kee-Hwan; Kim, Seung-Su; Hahn, Do-Hee

2008-01-01

The GIF(GEN IV International Forum) established an Economic Modelling Working Group(EMWG) in 2003 to create economic models and guidelines to facilitate in a future evaluation of the Generation IV nuclear energy systems and assess progress toward the GIF economic goals. These goals are to have a life cycle cost advantage over other energy sources, and to have a level of financial risk comparable to other energy projects. To do this, EMWG has been developed the G4-ECONS model, which is a generic EXCEL-based model for computation of the projected levelized unit electricity cost and/or levelized non-electricity unit product cost from GEN IV energy systems. KALIMER-600 has been developed as a new design concept based on the KALIMER-150 design. KALIMER-600 is a unique design concept which has a potential to achieve GEN IV technology goals even though there is a room for a design improvement in order to make the KALIMER-600 more competitive with future generation reactors. The objective of this study is to the assess economics of KALIMER-600 by using the G4-ECONS model

Preliminary Assessment of Transient of Over Power Accident for DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

TRACE code was selected as one of candidates for audit code, so sodium properties and heat transfer model in the code was verified first. On the basis of MARS-LMR code input, DSFR-600 TRACE model was developed and applied to PHTS tube rupture case, one of design base events (DBE) of DSFR-600. In this study, Transients of Over Power (TOP) event is assessed using TRACE code as one another case of DBEs of DSFR-600 for preparation of audit calculation of PGSFR.One of the design base events, transients of over power of Demonstration Sodium cooled Fast Reactor was simulated using TRACE code. Predicted fuel temperature showed that the peak fuel temperature occurs when the reactor scrammed and predicted temperature was similar to the MARS-LMRs assessment by KAERI. In this study, it is found that the second peak of fuel temperature is influenced by the inventory of steam generator and the natural circulation characteristic of the reactor vessel pool. Pre-calculation of the unprotected transients of over power with conservative reactivity assumption showed that this assumption is conservative in design base even assessment. However the method of measurement and applying the core radial, fuel and control rod axial expansion reactivity feedback is crucial in BDBE assessment of SFR.

Initial performance assessment of the Westinghouse AP600 containment design and related safety issues

International Nuclear Information System (INIS)

Nicolette, V.F.; Washington, K.E.; Tills, J.L.

1991-01-01

This work summarizes the Westinghouse AP600 advanced reactor design assessment calculations performed to date with the CONTAIN code. Correlations for modeling the important heat transfer phenomena are discussed as well. A CONTAIN model of the AP600 was constructed for design basis accident (DBA) calculations. Insights gained from modeling of the smaller-scale Westinghouse Integral Test Facility were incorporated in the development of the AP600 model. The results of the DBA calculations are compared to the results of other researchers to serve as a point of reference for future severe accident calculations. The CONTAIN calculations are reviewed to examine several parameters/phenomena of interest. The results of the calculations are also used to identify limitations of the CONTAIN code regarding application to advanced reactor containment designs. The most recent heat transfer correlations available in the literature are assessed for use in the flow regimes and geometries applicable to the AP600. Use of one of these correlations in CONTAIN may allow for a more accurate assessment of the AP600

MARS, 600 MWth NUCLEAR POWER PLANT

International Nuclear Information System (INIS)

Cumo, M.; Naviglio, A.; Sorabella, L.

2004-01-01

MARS (Multipurpose Advanced Reactor, inherently Safe) is a 600 MWth, single loop, pressurized light water reactor (PWR), developed at the Dept. of Nuclear Engineering and Energy Conversion of the University of Rome ''La Sapienza''. The design was focused to a multipurpose reactor to be used in high population density areas also for industrial heat production and, in particular, for water desalting. Using the well-proven technology and the operation experience of PWRs, the project introduces a lot of innovative features hugely improving the safety performance while keeping the cost of KWh competitive with traditional large power plants. Extensive use of passive safety, in depth plant simplification and decommissioning oriented design were the guidelines along the design development. The latest development in the plant design, in the decommissioning aspects and in the experimental activities supporting the project are shown in this paper

Dielectric Properties of Boron Nitride-Ethylene Glycol (BN-EG) Nanofluids

Science.gov (United States)

Fal, Jacek; Cholewa, Marian; Gizowska, Magdalena; Witek, Adam; ŻyŁa, GaweŁ

2017-02-01

This paper presents the results of experimental investigation of the dielectric properties of ethylene glycol (EG) with various load of boron nitride (BN) nanoparticles. The nanofuids were prepared by using a two-step method on the basis of commercially available BN nanoparticles. The measurements were carried out using the Concept 80 System (NOVOCONTROL Technologies GmbH & Co. KG, Montabaur, Germany) in a frequency range from 10 Hz to 10 MHz and temperatures from 278.15 K to 328.15 K. The frequency-dependent real (ɛ ^' }) and imaginary (ɛ ^' ' }) parts of the complex permittivity (ɛ ^*) and the alternating current (AC) conductivity are presented. Also, the effect of temperature and mass concentrations on the dielectric properties of BN-EG nanofluids are demonstrated. The results show that the most significant increase can be achieved for 20 wt.% of BN nanoparticles at 283.15 K and 288.15 K, that is eleven times larger than in the case of pure EG.

Fission and corrosion products behavior in primary circuits of LMFBR's

International Nuclear Information System (INIS)

Feuerstein, H.; Thorley, A.W.

1987-08-01

Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

Metallic behavior and enhanced adsorption energy of graphene on BN layer induced by Cu(111) substrate

International Nuclear Information System (INIS)

Hashmi, Arqum; Hong, Jisang

2014-01-01

We have investigated the adsorption properties and the electronic structure of graphene/BN and graphene/BN/Cu(111) systems by using van der Waals density functional theory. The ground-state adsorption site of graphene on BN/Cu(111) is found to be the same as that of graphene/BN. The Cu(111) substrate did not induce a significant change in the geometrical feature of graphene/BN. However, the adsorption energy of graphene on BN/Cu(111) is observed to be enhanced due to the Cu(111) substrate. In addition, we have found that the graphene layer displays a weak metallic character in graphene/BN/Cu(111) whereas an energy band gap is observed in the graphene in the graphene/BN bilayer system. Therefore, we have found that the metallic Cu(111) substrate affects the electronic structure and adsorption properties of graphene on BN/Cu(111), although it has no significant effect on the geometrical features.

Graphene/h-BN/GaAs sandwich diode as solar cell and photodetector.

Science.gov (United States)

Li, Xiaoqiang; Lin, Shisheng; Lin, Xing; Xu, Zhijuan; Wang, Peng; Zhang, Shengjiao; Zhong, Huikai; Xu, Wenli; Wu, Zhiqian; Fang, Wei

2016-01-11

In graphene/semiconductor heterojunction, the statistic charge transfer between graphene and semiconductor leads to decreased junction barrier height and limits the Fermi level tuning effect in graphene, which greatly affects the final performance of the device. In this work, we have designed a sandwich diode for solar cells and photodetectors through inserting 2D hexagonal boron nitride (h-BN) into graphene/GaAs heterostructure to suppress the static charge transfer. The barrier height of graphene/GaAs heterojunction can be increased from 0.88 eV to 1.02 eV by inserting h-BN. Based on the enhanced Fermi level tuning effect with interface h-BN, through adopting photo-induced doping into the device, power conversion efficiency (PCE) of 10.18% has been achieved for graphene/h-BN/GaAs compared with 8.63% of graphene/GaAs structure. The performance of graphene/h-BN/GaAs based photodetector is also improved with on/off ratio increased by one magnitude compared with graphene/GaAs structure.

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...

Influence of microstructure on stress corrosion cracking susceptibility of alloys 600 and 690 in primary water of pressurized water reactors

International Nuclear Information System (INIS)

Kergaravat, J.F.

1996-01-01

The mechanism(s) responsible for the stress corrosion cracking (SCC) of Alloy 600 steam generator tubes of pressurized water reactors remain misunderstood in spite of numerous studies on the subject. This failure mode presents several experimental similarities with intergranular creep fracture of austenitic stainless steels. As far as intergranular creep fracture is concerned, grain boundary sliding (GBS) was proved to favor failure. The aim of this work is to check the role played by GBS during SCC. It takes into account chemical (chromium content) and microstructural parameters (grain size, precipitation distribution and density). Therefore, to get a complete set of micro-structurally different samples, we have prepared solution annealed specimens (1100 deg C, 20 min., water quenched) from industrial tubes of Alloys 600 and 690. Each specimen was crept at 500 deg C (400 MPa), 430 deg C (425 MPa) and 360 deg C (475 MPa). Before testing, every sample were engraved with a 7 μm wide fiducial grid. This grid has allowed us to measure GBS after creep testing. GBS was observed for industrial and solution annealed samples for the three testing temperatures. GBS amplitude depends'on chromium content: for micro-structurally identical specimens, Alloy 600 exhibits more GB strain than Alloy 690. It also strongly depends on grain boundary precipitation characteristics: carbide free boundaries slide more easily. During in situ straining experiments performed in a transmission electronic microscope, GBS was evidenced at 320 deg C for Alloy 600 industrial samples. It consists in grain boundary dislocation motion in the interface plane. These dislocations originate from perfect dislocations gliding in the grain interior, encountering grain boundary and spreading in it. Metallic intergranular carbides provide strong obstacles to GBS so stress enhancements arise against them. These stress enhancements are released by micro-twin emission. Constant extension rate tensile tests were

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K.G.; Tretyakov, A.A.; Sorokin, Y.P.; Bondin, V.V.; Manakova, L.F.; Jardine, L.J.

2001-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is incineration

Sodium fires at fast reactors: RF status report

International Nuclear Information System (INIS)

Bagdasarov, Yu.E.; Buksha, Yu.K.; Drobyshev, A.V.; Zybin, V.A.; Ivanenko, V.N.; Kardash, D.Yu.; Kulikov, E.V.; Yagodkin, I.V.

1996-01-01

Scientific and engineering studies carried out in Russian Federation since 1992 up to 1996 in the sodium fire area and their main results are described. A review of activities on modification of the computer codes BOX and AERO developed at IPPE for calculating sodium fire consequences is given. Results of analysis of possible accidental situations at currently designed BN-800 reactor NPP with the use of these codes are presented. Sodium leaks occurring at our domestic fast reactors are briefly analyzed. Experimental work performed are described. Results of comparative analysis of common-cause and sodium fire hazards for fast reactor NPP are presented. (author)

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

44Sc-DOTA-BN[2-14]NH2 in comparison to 68Ga-DOTA-BN[2-14]NH2 in pre-clinical investigation. Is 44Sc a potential radionuclide for PET?

International Nuclear Information System (INIS)

Koumarianou, E.; Loktionova, N.S.; Fellner, M.; Roesch, F.; Thews, O.; Pawlak, D.; Archimandritis, S.C.; Mikolajczak, R.

2012-01-01

Aim: In the present study we demonstrate the in vitro and in vivo comparison of the 44 Sc and 68 Ga labeled DOTA-BN[2-14]NH 2 . 44 Sc is a positron emitter with a half life of 3.92 h. Hence it could be used for PET imaging with ligands requiring longer observation time than in the case of 68 Ga. Methods: The binding affinity of nat Sc-DOTA-BN[2-14]NH 2 and nat Ga-DOTA-BN[2-14]NH 2 to GRP receptors was studied in competition to [ 125 I-Tyr 4 ]-Bombesin in the human prostate cancer cell line PC-3. A preliminary biodistribution in normal rats was performed, while first microPET images were assessed in male Copenhagen rats bearing the androgen-independent Dunning R-3327-AT-1 prostate cancer tumor. Results: The affinity to GRP receptors in the PC-3 cell line was higher for nat Ga-DOTA-BN[2-14]NH 2 (IC 50 (nM)=0.85±0.06) than that of nat Sc-DOTA-BN[2-14]NH 2 (IC 50 (nM)=6.49±0.13). The internalization rate of 68 Ga labeled DOTA-BN[2-14]NH 2 was slower than that of 44 Sc, but their final internalization percents were comparable. 68 Ga-DOTA-BN[2-14]NH 2 was externalized faster than 44 Sc-DOTA-BN[2-14]NH 2 . The biodistribution of 44 Sc-DOTA-BN[2-14]NH 2 and 68 Ga-DOTA-BN[2-14]NH 2 in normal rats revealed a higher uptake in target organs and tissues of the first one while both excreted mainly through urinary tract. In microPET images both tracers were accumulated in the tumor with similar uptake patterns. Conclusions: Despite the differences in the receptor affinity both the 68 Ga- and the 44 Sc-labeled DOTA-BN[2-14]NH 2 tracers showed comparable distribution and similar time constants of uptake and elimination. Moreover no differences in tumor accumulation (neither in the overall uptake nor in the dynamics) were observed from the microPet imaging. From that perspective the use of either 44 Sc or 68 Ga for detecting tumors with GRP receptors is equivalent. - Highlights: ► In vitro and in vivo evaluation of 44 Sc- and 68 Ga-DOTA-BN[2-14]NH 2 in reference to published

Method for calculating the forces and deformations in the fast reactor fuel assembly accounting for the effects of reactor control system elements and shutdown

International Nuclear Information System (INIS)

Likhachev, Yu.I.; Vashlyaev, Yu.N.; Kravchenko, I.N.

1980-01-01

Methods for calculating deformations and interaction forces of heat-generating assemblies (HGA) of fast reactor core with account for the effect of control and protection system (CPS) elements at the reactor operation and change of interaction efforts between HGA at the reactor shutdown, are described. The results of testing the suggested methods on example of estimate of HGA behaviour of the BN-350 reactor are presented. For estimating the effect of CPS elements on HGA bending the sector model has been used. It is assumed that HGA deformation inside each sector is independent of HGA deformation of other sectors. A higher calculation accuracy is attained by means of laying out of sectors into regions of preferable influence of emergency protection elements and compensating packets. When determining deformation and interaction efforts between HGA caused by temperature change in the course of shutdown it is supposed that the HGA deformation is purely elastic. The methods described are realized in the form of ABRI-CPS and ABRI-HOL programs written in FORTRAN for the BESM-6 computer. The results of HGA calculations of the BN-350 reactor core show that CPS elements decrease contact efforts in the middle of the central packet, increase contact efforts in the peak of the central packet, increase contact efforts in the peaks of packets from the eight row to the periphery and increase contact efforts in the middles of packets from the 5th to 9th row [ru

Flattening and manipulation of the electronic structure of h-BN/Rh(111) nanomesh upon Sn intercalation

Science.gov (United States)

Sugiyama, Yuya; Bernard, Carlo; Okuyama, Yuma; Ideta, Shin-ichiro; Tanaka, Kiyohisa; Greber, Thomas; Hirahara, Toru

2018-06-01

We have deposited Sn on corrugated hexagonal boron nitride (h-BN) nanomeshs formed on Rh(111) and found that Sn atoms are intercalated between h-BN and Rh, flattening the h-BN. Our reflection high-energy electron diffraction (RHEED) analysis showed that the average in-plane lattice constant of h-BN increases due to the loss of the corrugation. Furthermore, electronic structure measurements based on angle-resolved photoemission spectroscopy (ARPES) showed that the h-BN π band width increases significantly while the σ band width does not change as much. These behaviors were partly different from previous reports on the intercalation of h-BN/Rh system. Our results offer a novel, simple method to control the electronic structure of h-BN.

On the Difference Equation xn=anxn-k/(bn+cnxn-1⋯xn-k

Directory of Open Access Journals (Sweden)

Stevo Stević

2012-01-01

Full Text Available The behavior of well-defined solutions of the difference equation xn=anxn-k/(bn+cnxn-1⋯xn-k, n∈ℕ0, where k∈ℕ is fixed, the sequences an, bn and cn are real, (bn,cn≠(0,0, n∈ℕ0, and the initial values x-k,…,x-1 are real numbers, is described.

Testing of the dual slab verification detector for attended measurements of the BN-350 dry storage casks

Energy Technology Data Exchange (ETDEWEB)

Santi, Peter A [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Williams, Richard B [Los Alamos National Laboratory; Parker, Robert F [Los Alamos National Laboratory

2009-01-01

The Dual Slab Verification Detector (DSVD) has been developed and built by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of {sup 3}He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. The DSVD will be used to perform measurements of the neutron flux emanating from inside the dry storage cask at several locations around each cask to establish a neutron 'fingerprint' that is sensitive to the contents of the cask. The sensitivity of the fingerprinting technique to the removal of specific amount of nuclear material from the cask is determined by the characteristics of the detector that is used to perform the measurements, the characteristics of the spent fuel being measured, and systematic uncertainties that are associated with the dry storage scenario. MCNPX calculations of the BN-350 dry storage asks and layout have shown that the neutron fingerprint verification technique using measurements from the DSVD would be sensitive to both the amount and location of material that is present within an individual cask. To confirm the performance of the neutron fingerprint technique in verifying the presence of BN-350 spent fuel in dry storage, an initial series of measurements have been performed to test the performance and characteristics of the DSVD. Results of these measurements will be presented and compared with MCNPX results.

Biological Efficacy of Streptomyces sp. Strain BN1 against the Cereal Head Blight Pathogen Fusarium graminearum

Directory of Open Access Journals (Sweden)

Boknam Jung

2013-03-01

Full Text Available Fusarium head blight (FHB caused by the filamentous fungus Fusarium graminearum is one of the most severe diseases threatening the production of small grains. Infected grains are often contaminated with mycotoxins such as zearalenone and trichothecences. During survey of contamination by FHB in rice grains, we found a bacterial isolate, designated as BN1, antagonistic to F. graminearum. The strain BN1 had branching vegetative hyphae and spores, and its aerial hyphae often had long, straight filaments bearing spores. The 16S rRNA gene of BN1 had 100% sequence identity with those found in several Streptomyces species. Phylogenetic analysis of ITS regions showed that BN1 grouped with S. sampsonii with 77% bootstrap value, suggesting that BN1 was not a known Streptomyces species. In addition, the efficacy of the BN1 strain against F. graminearum strains was tested both in vitro and in vivo. Wheat seedling length was significantly decreased by F. graminearum infection. However, this effect was mitigated when wheat seeds were treated with BN1 spore suspension prior to F. graminearum infection. BN1 also significantly decreased FHB severity when it was sprayed onto wheat heads, whereas BN1 was not effective when wheat heads were point inoculated. These results suggest that spraying of BN1 spores onto wheat heads during the wheat flowering season can be efficient for plant protection. Mechanistic studies on the antagonistic effect of BN1 against F. graminearum remain to be analyzed.

44Sc-DOTA-BN[2-14]NH2 in comparison to 68Ga-DOTA-BN[2-14]NH2 in pre-clinical investigation. Is 44Sc a potential radionuclide for PET?

Science.gov (United States)

Koumarianou, E; Loktionova, N S; Fellner, M; Roesch, F; Thews, O; Pawlak, D; Archimandritis, S C; Mikolajczak, R

2012-12-01

In the present study we demonstrate the in vitro and in vivo comparison of the (44)Sc and (68)Ga labeled DOTA-BN[2-14]NH(2). (44)Sc is a positron emitter with a half life of 3.92 h. Hence it could be used for PET imaging with ligands requiring longer observation time than in the case of (68)Ga. The binding affinity of (nat)Sc-DOTA-BN[2-14]NH(2) and (nat)Ga-DOTA-BN[2-14]NH(2) to GRP receptors was studied in competition to [(125)I-Tyr(4)]-Bombesin in the human prostate cancer cell line PC-3. A preliminary biodistribution in normal rats was performed, while first microPET images were assessed in male Copenhagen rats bearing the androgen-independent Dunning R-3327-AT-1 prostate cancer tumor. The affinity to GRP receptors in the PC-3 cell line was higher for (nat)Ga-DOTA-BN[2-14]NH(2) (IC(50)(nM)=0.85 ± 0.06) than that of (nat)Sc-DOTA-BN[2-14]NH(2) (IC(50) (nM)=6.49 ± 0.13). The internalization rate of (68)Ga labeled DOTA-BN[2-14]NH(2) was slower than that of (44)Sc, but their final internalization percents were comparable. (68)Ga-DOTA-BN[2-14]NH(2) was externalized faster than (44)Sc-DOTA-BN[2-14]NH(2). The biodistribution of (44)Sc-DOTA-BN[2-14]NH(2) and (68)Ga-DOTA-BN[2-14]NH(2) in normal rats revealed a higher uptake in target organs and tissues of the first one while both excreted mainly through urinary tract. In microPET images both tracers were accumulated in the tumor with similar uptake patterns. Despite the differences in the receptor affinity both the (68)Ga- and the (44)Sc-labeled DOTA-BN[2-14]NH(2) tracers showed comparable distribution and similar time constants of uptake and elimination. Moreover no differences in tumor accumulation (neither in the overall uptake nor in the dynamics) were observed from the microPet imaging. From that perspective the use of either (44)Sc or (68)Ga for detecting tumors with GRP receptors is equivalent. Copyright © 2012 Elsevier Ltd. All rights reserved.

Characteristics of fast reactor core designs and closed fuel cycle

International Nuclear Information System (INIS)

Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

2007-01-01

On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

Process heat applications of HTR-PM600 in Chinese petrochemical industry: Preliminary study of adaptability and economy

International Nuclear Information System (INIS)

Fang, Chao; Min, Qi; Yang, Yanran; Sun, Yuliang

2017-01-01

Highlights: •High Temperature Gas Cooled Reactor (HTGR) could work as heat source for petrochemical industry. •The joint of a 600 MW modular HTGR (HTR-PM600) and petrochemical industry is achievable. •The mature technology of turbine in thermal power station could be readily adopted. •The economy of this scheme is also acceptable. -- Abstract: High Temperature Gas Cooled Reactor (HTGR) could work as heat source for petrochemical industry. In this article, the preliminary feasibility of a 600 MW modular HTGR (HTR-PM600) working as heat source for a typical hypothetical Chinese petrochemical factory is discussed and it is found that the joint of HTR-PM600 and petrochemical industry is achievable. In detail, the heat and water balance analysis of the petrochemical factory is given. Furthermore, the direct cost of heat supplied by HTR-PM600 is calculated and corresponding economy is estimated. The results show that though there are several challenges, the application of process heat of HTGR to petrochemical industry is practical in sense of both technology and economy.

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

International Nuclear Information System (INIS)

Bruschi, H.J.

1996-01-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse's advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

Energy Technology Data Exchange (ETDEWEB)

Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1996-10-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse`s advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

Decolourization and degradation of azo Dye, Synozol Red HF6BN ...

African Journals Online (AJOL)

Decolourization and degradation of azo Dye, Synozol Red HF6BN, by Pleurotus ostreatus. Sidra Ilyas, Skinder Sultan Sultan, Abdul Rehman. Abstract. The present paper focuses on the use of fungus, Pleurotus ostreatus, to decolorize and degrade azo dye, Synazol Red HF6BN. Decolorization study showed that P.

AM600: A new look at the nuclear steam cycle

International Nuclear Information System (INIS)

Field, Robert M.

2017-01-01

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections

AM600: A new look at the nuclear steam cycle

Energy Technology Data Exchange (ETDEWEB)

Field, Robert M. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2017-04-15

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.

AM600: A New Look at the Nuclear Steam Cycle

Directory of Open Access Journals (Sweden)

Robert M. Field

2017-04-01

Full Text Available Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.

Enhancing the mechanical properties of BN nanosheet-polymer composites by uniaxial drawing

Science.gov (United States)

Jan, Rahim; May, Peter; Bell, Alan P.; Habib, Amir; Khan, Umar; Coleman, Jonathan N.

2014-04-01

We have used liquid exfoliation of hexagonal Boron-Nitride (BN) to prepare composites of BN nanosheets of three different sizes in polyvinylchloride matrices. These composites show low levels of reinforcement, consistent with poor alignment of the nanosheets as-described by a modified version of Halpin-Tsai theory. However, drawing of the composites to 300% strain results in a considerable increase in mechanical properties with the maximum composite modulus and strength both ~×3 higher than that of the pristine polymer. In addition, the rate of increase of modulus with BN volume fraction was up to 3-fold larger than for the unstrained composites. This is higher than can be explained by drawing-induced alignment using Halpin-Tsai theory. However, the data was consistent with a combination of alignment and strain-induced de-aggregation of BN multilayers.

Development of a three dimensional homogeneous calculation model for the BFS-62 critical experiment. Preparation of adjusted equivalent measured values for sodium void reactivity values. Final report

International Nuclear Information System (INIS)

Manturov, G.; Semenov, M.; Seregin, A.; Lykova, L.

2004-01-01

The BFS-62 critical experiments are currently used as 'benchmark' for verification of IPPE codes and nuclear data, which have been used in the study of loading a significant amount of Pu in fast reactors. The BFS-62 experiments have been performed at BFS-2 critical facility of IPPE (Obninsk). The experimental program has been arranged in such a way that the effect of replacement of uranium dioxied blanket by the steel reflector as well as the effect of replacing UOX by MOX on the main characteristics of the reactor model was studied. Wide experimental program, including measurements of the criticality-keff, spectral indices, radial and axial fission rate distributions, control rod mock-up worth, sodium void reactivity effect SVRE and some other important nuclear physics parameters, was fulfilled in the core. Series of 4 BFS-62 critical assemblies have been designed for studying the changes in BN-600 reactor physics from existing state to hybrid core. All the assemblies are modeling the reactor state prior to refueling, i.e. with all control rod mock-ups withdrawn from the core. The following items are chosen for the analysis in this report: Description of the critical assembly BFS-62-3A as the 3rd assembly in a series of 4 BFS critical assemblies studying BN-600 reactor with MOX-UOX hybrid zone and steel reflector; Development of a 3D homogeneous calculation model for the BFS-62-3A critical experiment as the mock-up of BN-600 reactor with hybrid zone and steel reflector; Evaluation of measured nuclear physics parameters keff and SVRE (sodium void reactivity effect); Preparation of adjusted equivalent measured values for keff and SVRE. Main series of calculations are performed using 3D HEX-Z diffusion code TRIGEX in 26 groups, with the ABBN-93 cross-section set. In addition, precise calculations are made, in 299 groups and Ps-approximation in scattering, by Monte-Carlo code MMKKENO and discrete ordinate code TWODANT. All calculations are based on the common system

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

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

Anticorrosive performance of waterborne epoxy coatings containing water-dispersible hexagonal boron nitride (h-BN) nanosheets

Science.gov (United States)

Cui, Mingjun; Ren, Siming; Chen, Jia; Liu, Shuan; Zhang, Guangan; Zhao, Haichao; Wang, Liping; Xue, Qunji

2017-03-01

Homogenous dispersion of hexagonal boron nitride (h-BN) nanosheets in solvents or in the polymer matrix is crucial to initiate their many applications. Here, homogeneous dispersion of hexagonal boron nitride (h-BN) in epoxy matrix was achieved with a water-soluble carboxylated aniline trimer derivative (CAT-) as a dispersant, which was attributed to the strong π-π interaction between h-BN and CAT-, as proved by Raman and UV-vis spectra. Transmission electron microscopy (TEM) analysis confirmed a random dispersion of h-BN nanosheets in the waterborne epoxy coatings. The deterioration process of water-borne epoxy coating with and without h-BN nanosheets during the long-term immersion in 3.5 wt% NaCl solution was investigated by electrochemical measurements and water absorption test. Results implied that the introduction of well dispersed h-BN nanosheets into waterborne epoxy system remarkably improved the corrosion protection performance to substrate. Moreover, 1 wt% BN/EP composite coated substrate exhibited higher impedance modulus (1.3 × 106 Ω cm2) and lower water absorption (4%) than those of pure waterborne epoxy coating coated electrode after long-term immersion in 3.5 wt% NaCl solution, demonstrating its superior anticorrosive performance. This enhanced anticorrosive performance was mainly ascribed to the improved water barrier property of epoxy coating via incorporating homogeneously dispersed h-BN nanosheets.

Double vacancy on BN layer: A natural trap for Hydrogen Molecule

International Nuclear Information System (INIS)

Arellano, J S

2015-01-01

A pair of vacancies, one of boron and other of nitrogen atom at a flat layer becomes a natural trap to capture a hydrogen molecule at the center of the cavity defined by the empty space left by the lack of a nitrogen and a boron atom at the perfect BN layer formed by 16 N atoms and 16 B atoms. The adsorption of the hydrogen molecule is compared with the equivalent graphene layer with a pair of carbon vacancies. The little increase in the BN cell parameter respect to the graphene cell parameter, besides the differences between N, B and C atoms helps to explain the easier adsorption on the defective BN layer

Current limiting experiment with 600 V/100A rectification type superconducting fault current limiter; 600 V-100A kyu seiryugata chodendo genryuki no genryu shiken

Energy Technology Data Exchange (ETDEWEB)

Matsuzaki, J.; Tsurunaga, K.; Urata, M. [Toshiba Corp., Tokyo (Japan); Okuma, T.; Sato, Y.; Iwata, Y. [Tokyo Electric Power Co., Inc., Tokyo (Japan)

1999-06-07

The rectification type current limiter with the current-limiting system of the new type which combined rectifier circuits with the direct current reactor has been proposed until now, and it has succeeded in the current-limiting test by the normal conduction reactor by the 6.6kV class model vessel. Since the loss of the conductor becomes fundamentally the zero, in the same current limiter, by using superconducting wire rod, because direct current always flows in the reactor, making into low-loss becomes possible. In this report, this paper describes cut-off characteristic of 600V/100A rectification type superconductive current limiter using the metal type superconductive conductor. (NEDO)

Modulation of band gap by an applied electric field in BN-based heterostructures

Science.gov (United States)

Luo, M.; Xu, Y. E.; Zhang, Q. X.

2018-05-01

First-principles density functional theory (DFT) calculations are performed on the structural and electronic properties of the SiC/BN van der Waals (vdW) heterostructures under an external electric field (E-field). Our results reveal that the SiC/BN vdW heterostructure has a direct band gap of 2.41 eV in the raw. The results also imply that electrons are likely to transfer from BN to SiC monolayer due to the deeper potential of BN monolayer. It is also observed that, by applying an E-field, ranging from -0.50 to +0.65 V/Å, the band gap decreases from 2.41 eV to zero, which presents a parabola-like relationship around 0.0 V/Å. Through partial density of states (PDOS) plots, it is revealed that, p orbital of Si, C, B, and N atoms are responsible for the significant variations of band gap. These obtained results predict that, the electric field tunable band gap of the SiC/BN vdW heterostructures carries potential applications for nanoelectronics and spintronic device applications.

Power ramp tests of BWR-MOX fuels

International Nuclear Information System (INIS)

Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

1996-01-01

Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

RESOLVING THE STRUCTURE AND KINEMATICS OF THE BN OBJECT AT 0.''2 RESOLUTION

International Nuclear Information System (INIS)

Rodriguez, Luis F.; Zapata, Luis A.; Ho, Paul T. P.

2009-01-01

We present sensitive 7 mm observations of the H53α recombination line and adjacent continuum, made toward the Orion BN/KL region. In the continuum we detect the BN object, the radio source I (GMR I) and the radio counterpart of the infrared (IR) source n (Orion-n). Comparing with observations made at similar angular resolutions but lower frequency, we discuss the spectral indices and angular sizes of these sources. In the H53α line, we only detect the BN object. This is the first time that radio recombination lines have been detected from this source. The LSR radial velocity of BN from the H53α line, v LSR = 20.1 ± 2.1 km s -1 , is consistent with that found from previous studies in near-IR lines. While the continuum emission is expected to have considerable optical depth at 7 mm, the observed H53α line emission is consistent with an optically thin nature and we discuss possible explanations for this apparent discrepancy. There is evidence of a velocity gradient, with the NE part of BN being redshifted by ∼10 km s -1 with respect to the SW part. This is consistent with the suggestion of Jiang et al. that BN may be driving an ionized outflow along that direction.

Thermal Aging Effects on Heat Affected Zone of Alloy 600 in Dissimilar Metal Weld

Energy Technology Data Exchange (ETDEWEB)

Ham, Jun Hyuk; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-05-15

Dissimilar metal weld (DMW), consists of Alloy 600, Alloy 182, and A508 Gr.3, is now being widely used as the reactor pressure vessel penetration nozzle and the steam generator tubing material for pressurized water reactors (PWR) because of its mechanical property, thermal expansion coefficient, and corrosion resistance. The heat affected zone (HAZ) on Alloy 600 which is formed by welding process is critical to crack. According to G.A. Young et al. crack growth rates (CGR) in the Alloy 600 HAZ were about 30 times faster than those in the Alloy 600 base metal tested under the same conditions [3]. And according to Z.P. Lu et al. CGR in the Alloy 600 HAZ can be more than 20 times higher than that in its base metal. To predict the life time of components, there is a model which can calculate the effective degradation years (EDYs) of the material as a function of operating temperature. This study was conducted to investigate how thermal aging affects the hardness of dissimilar metal weld from the fusion boundary to Alloy 600 base metal and the residual strain at Alloy 600 heat affected zone. Following conclusions can be drawn from this study. The hardness, measured by Vickers hardness tester, peaked near the fusion boundary between Alloy 182 and Alloy 600, and it decreases as the picked point goes to Alloy 600 base metal. Even though the formation of precipitate such as Cr carbide, thermal aging doesn't affect the value and the tendency of hardness because of reduced residual stress. According to kernel average misorientation mapping, residual strain decreases when the material thermally aged. And finally, in 30 years simulated specimen, the high residual strain almost disappears. Therefore, the influence of residual strain on primary water stress corrosion cracking can be diminished when the material undergoes thermal aging.

Thermal Conductivity of Epoxy Resin Composites Filled with Combustion Synthesized h-BN Particles.

Science.gov (United States)

Chung, Shyan-Lung; Lin, Jeng-Shung

2016-05-20

The thermal conductivity of epoxy resin composites filled with combustion-synthesized hexagonal boron nitride (h-BN) particles was investigated. The mixing of the composite constituents was carried out by either a dry method (involving no use of solvent) for low filler loadings or a solvent method (using acetone as solvent) for higher filler loadings. It was found that surface treatment of the h-BN particles using the silane 3-glycidoxypropyltrimethoxysilane (GPTMS) increases the thermal conductivity of the resultant composites in a lesser amount compared to the values reported by other studies. This was explained by the fact that the combustion synthesized h-BN particles contain less -OH or active sites on the surface, thus adsorbing less amounts of GPTMS. However, the thermal conductivity of the composites filled with the combustion synthesized h-BN was found to be comparable to that with commercially available h-BN reported in other studies. The thermal conductivity of the composites was found to be higher when larger h-BN particles were used. The thermal conductivity was also found to increase with increasing filler content to a maximum and then begin to decrease with further increases in this content. In addition to the effect of higher porosity at higher filler contents, more horizontally oriented h-BN particles formed at higher filler loadings (perhaps due to pressing during formation of the composites) were suggested to be a factor causing this decrease of the thermal conductivity. The measured thermal conductivities were compared to theoretical predictions based on the Nielsen and Lewis theory. The theoretical predictions were found to be lower than the experimental values at low filler contents ( 60 vol %).

Ab initio study of the structural, electronic and optical properties of BAs and BN compounds and BN{sub x}As{sub 1−x} alloys

Energy Technology Data Exchange (ETDEWEB)

Guemou, M., E-mail: guemoumhamed7@gmail.com [Engineering Physics Laboratory, Ibn Khaldoun University of Tiaret, Postbox 78-Zaaroura, 14000 Tiaret (Algeria); Abdiche, A.; Riane, R. [Applied Materials Laboratory, Research Center, University of Sidi Bel Abbes, 22000 Sidi Bel Abbes (Algeria); Khenata, R. [Laboratoire de Physique Quantique et de Modélisation Mathématique (LPQ3M), Département de Technologie, Université de Mascara, 29000 Mascara (Algeria)

2014-03-01

In this work, we present a density-functional theory study of structural, electronic and optical properties of BAs, BN binary compounds and their ternary BN{sub x}As{sub 1−x} solid solutions. The calculations are done by using the all-electron full potential linear augmented plane-wave method (FP-LAPW) as employed in WIEN2k code. For the exchange-correlation potential, local-density approximation (LDA) and generalized gradient approximation (GGA) have been used to calculate theoretical lattice parameters, bulk modulus, and its pressure derivative. The electronic band structure of these compounds have been calculated by using the above two approximations. We have also investigated in this article the density of state and the optical properties such as the dielectric function and the refractive index of BAs, BN and BN{sub 0.25}As{sub 0.75} compounds by using the above method. The results obtained for structural and electronic properties are compared with experimental data and other computational work. It has been found that the energy bands with all these approximations are similar except the band gap values. It has also been found that our results with LDA and GGA are in good agreement with other computational work wherever these are available.

Cubic boron nitride coatings for innovative applications; Schichten aus kubischem Bornitrid (cBN) fuer innovative Anwendungen

Energy Technology Data Exchange (ETDEWEB)

Keunecke, M.; Bewilogua, K. [Fraunhofer Inst. fuer Schicht- und Oberflaechentechnik (Germany)

2001-08-01

Coatings of cubic boron nitride (cBN), the second hardest of all known materials, were prepared using a sputter process. A new coating design, based on a special B-C-N layer system, allows to deposit thick (> 2 {mu}m) cBN films, however so far only on silicon substrates. The properties of the coatings are quite similar to those of bulk cBN. Promising experiments were performed with respect to a transfer of this application relevant layer system to cemented carbide and steel substrates. First measurements of the mechanical and tribological properties confirmed the outstanding properties and the high potential of such cBN based coating systems. (orig.) [German] Schichten aus kubischem Bornitrid (cBN), dem nach Diamant zweithaertesten aller bekannten Materialien, wurden mit einem Sputter-Prozess hergestellt. Ein neuartiger Schichtaufbau, der auf einem speziellen B-C-N-Schichtsystem basiert, ermoeglicht die Abscheidung von cBN-Schichten mit ueber 2 {mu}m Dicke, allerdings bisher nur auf Siliciumsubstraten. Die Eigenschaften der Schichten sind denen von massivem cBN sehr aehnlich. Es wurden vielversprechende Experimente zur Uebertragung dieses fuer Werkzeugbeschichtungen und vielfaeltige andere Anwendungen interessanten Schichtsystems auf Werkzeugsubstrate durchgefuehrt. Erste Untersuchungen der mechanisch-tribologischen Eigenschaften der auf Hartmetall- und Stahlsubstraten abgeschiedenen Schichten belegen das aussergewoehnlich hohe Potential der cBN-basierten Schichtsysteme. (orig.)

Diamond and cBN hybrid and nanomodified cutting tools with enhanced performance: Development, testing and modelling

DEFF Research Database (Denmark)

Loginov, Pavel; Mishnaevsky, Leon; Levashov, Evgeny

2015-01-01

with 25% of diamond replaced by cBN grains demonstrate 20% increased performance as compared with pure diamond machining tools, and more than two times higher performance as compared with pure cBN tools. Further, cast iron machining efficiency of the wheels modified by hBN particles was 80% more efficient......The potential of enhancement of superhard steel and cast iron cutting tool performance on the basis of microstuctural modifications of the tool materials is studied. Hybrid machining tools with mixed diamond and cBN grains, as well as machining tool with composite nanomodified metallic binder...... are developed, and tested experimentally and numerically. It is demonstrated that both combination of diamond and cBN (hybrid structure) and nanomodification of metallic binder (with hexagonal boron nitride/hBN platelets) lead to sufficient improvement of the cast iron machining performance. The superhard tools...

Oxidation Behavior of AlN/h-BN Nano Composites at High Temperature

International Nuclear Information System (INIS)

Jin Haiyun; Huang Yinmao; Feng Dawei; He Bo; Yang Jianfeng

2011-01-01

Both AlN/ nano h-BN composites and AlN/ micro h-BN composites were fabricated. The high temperature oxidation behaviors were investigated at 1000deg. C and 1300deg. C using a cycle-oxidation method. The results showed that there were little changes of both nano composites and monolithic AlN ceramic at temperature of 1000deg. C. And at 1300deg. C, the oxidation dynamics curve of composites could be divided into two courses: a slowly weight increase and a rapid weight decrease, but the oxidation behavior of nano composites was better than micro composites. It was due to that the uniform distribution of oxidation production (Al 18 B 4 O 33 ) surround the AlN grains in nano composites and the oxidation proceeding was retarded. The XRD analysis and SEM observations showed that there was no BN remained in the composites surface after 1300deg. C oxidation and the micropores remain due to the vaporizing of B 2 O 3 oxidized by BN.

Technical feasibility and reliability of passive safety systems of AC600

International Nuclear Information System (INIS)

Niu, W.; Zeng, X.

1996-01-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished by the Nuclear Power Institute of China. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also

Technical feasibility and reliability of passive safety systems of AC600

Energy Technology Data Exchange (ETDEWEB)

Niu, W; Zeng, X [Nuclear Power Inst. of China, Chendu (China)

1996-12-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also involved. (author). 3 figs, 1 tab.

Research on Operation and Control Strategy of 600MW PWR in Load Follow

Energy Technology Data Exchange (ETDEWEB)

Qu, Bing Yang; Cao, Xin Rong [Harbin Engineering University, Harbin (China); Li, Han Chen [China Nuclear Power Engineering Co., Beijing (China)

2014-08-15

600MW Pressurized Water Reactor (PWR) is designed to operate in Constant Axial Offset Control (CAOC) strategy with base load originally. By calculations over a typical load follow scenario '12-3-6-3 {sup (}100-50-100%FP) via the CASMO-4E and SIMULATE-3 package, values of core operating parameter have been examined. With the progress of the nuclear power industry, advanced reactors are considered to have a good performance in load follow, economy and flexibility. Under the premise of fuel loading and structural dimensions unchanged, two independent control rod groups M and AO are used in 600MW pressurized water reactor to provide fine control of both the core reactivity and axial power distribution, which is named ' Improved G strategy .' The influences of different control rod distributions, composition materials, and overlap steps had in power changes have been examined in a comparative study to choose the optimal one.Then we simulate a range of load follow scenarios of the redesigned 600MW core without adjusting soluble boron concentration in the begin, middle and end of first cycle. This paper additionally demonstrated the moderator temperature coefficient and shutdown margin values of the reactor in Improved G strategy to compare with the thermal safety design criteria. It's demonstrated that adequate adjustment of control rod groups enable the core to perform load follow through Improved G strategy in 80% of cycle and save a large volume of liquid effluent particularly toward the end of cycle.

Code HEX-Z-DMG for support of accounting for and control of nuclear material software system as part of international safeguards system at BN-350 site

International Nuclear Information System (INIS)

Bushmakin, A.G.; Schaefer, B.

1999-01-01

A code for the computation of the global neutron distribution in the three-dimensional hexagonal-z geometry and multi-group diffusion approximation was developed at BN-350 as the main part of the BN-350 accounting for and control of nuclear material software system. This software system includes: the model for stationary distributions of neutrons; the model to calculate isotope compositions changing; the model of refueling operations; To develop this system next two principal problems were solved: to make a micro cross sections library for all nuclides for the BN-350 reactor core; to develop the code for the computation of the global neutron distribution. To solve first task the twenty-six-energy-groups micro cross sections library for more than seventy nuclides was produced. To solve second task the three-dimensional hexagonal-z geometry and multi-group diffusion approximation code was developed. This code (HEX-Z-DMG) was based on the solution of the multi groups diffusion equation using the standard net approach. The series of calculations was performed in the twenty-six-energy-groups representation using this code. We compared eigenvalues (k eff ), a worth added during refueling operations, spatial and energy-group-dependent neutron flux distributions with results of calculation using other code (DIF3D). After the series of these calculations we can say that the HEX-Z-DMG code is well established to use as the part of the BN-350 accounting for and control of nuclear material software system. (author)

Effect of Ti3+ ion on the Corrosion Behavior of Alloy 600

International Nuclear Information System (INIS)

Lee, Chang Bong; Lim, Han Gwi; Kim, Bok Hee; Kim, Ki Ju

1999-01-01

Alloy 600 has been widely used as a steam generator tubing material in pressurized water reactors(PWRs) nuclear power plants. Corrosion of steam generator tubing mainly occurs on the secondary water side. The purpose of this work is primarily concerned with examining the effect of Ti 3+ ion concentrations on the corrosion behavior of the Alloy 600 steam generator tubing material. Corrosion behavior of the Alloy 600 steam generator tubing material was studied in aqueous solutions with varying Ti 3+ ion concentration at room temperature. Potentiodynamic and potentiostatic polarization techniques were used to determine the corrosion and pitting potentials for the Alloy 600 test material. The addition of Ti 3+ ion to 1000ppm, showed inhibition effect on the corrosion of Alloy 600. But the corrosion of Alloy 600 was accelerated when the concentration of Ti 3+ ion exceeded 1000ppm, it is assumed that the effect of general corrosion of Alloy 600 is more sensitive than pitting corrosion. It is considered that the passive film which was formed on the Alloy 600 surface in the 100ppm Ti 3+ ion containing solution is mainly consisted of TiO 2

Midgap states and band gap modification in defective graphene/h-BN heterostructures

NARCIS (Netherlands)

Sachs, B.; Wehling, T.O.; Katsnelson, M.I.; Lichtenstein, A.I.

2016-01-01

The role of defects in van der Waals heterostructures made of graphene and hexagonal boron nitride (h-BN) is studied using a combination of ab initio and model calculations. Despite the weak van der Waals interaction between layers, defects residing in h-BN, such as carbon impurities and antisite

Anticorrosive performance of waterborne epoxy coatings containing water-dispersible hexagonal boron nitride (h-BN) nanosheets

Energy Technology Data Exchange (ETDEWEB)

Cui, Mingjun [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); University of Chinese Academy of Sciences, Beijing 100039 (China); Ren, Siming [Key Laboratory of Marine Materials and Related Technologies, Zhejiang Key Laboratory of Marine Materials and Protective Technologies, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo 315201 (China); University of Chinese Academy of Sciences, Beijing 100039 (China); Chen, Jia; Liu, Shuan [Key Laboratory of Marine Materials and Related Technologies, Zhejiang Key Laboratory of Marine Materials and Protective Technologies, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo 315201 (China); Zhang, Guangan [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Zhao, Haichao, E-mail: zhaohaichao@nimte.ac.cn [Key Laboratory of Marine Materials and Related Technologies, Zhejiang Key Laboratory of Marine Materials and Protective Technologies, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo 315201 (China); Wang, Liping, E-mail: wangliping@nimte.ac.cn [Key Laboratory of Marine Materials and Related Technologies, Zhejiang Key Laboratory of Marine Materials and Protective Technologies, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo 315201 (China); Xue, Qunji, E-mail: qjxue@lzb.ac.cn [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Key Laboratory of Marine Materials and Related Technologies, Zhejiang Key Laboratory of Marine Materials and Protective Technologies, Ningbo Institute of Materials Technology and Engineering, Chinese Academy of Sciences, Ningbo 315201 (China)

2017-03-01

Highlights: • Hexagonal boron nitride nanosheets were well dispersed by using water-soluble carboxylated aniline trimer as dispersant. • The best corrosion performance of waterborne epoxy coatings was achieved with the addition of 1 wt% h-BN. • The decrease of the pores and defects of coating matrix inhibits the diffusion and water absorption of corrosive medium in the coating. - Abstract: Homogenous dispersion of hexagonal boron nitride (h-BN) nanosheets in solvents or in the polymer matrix is crucial to initiate their many applications. Here, homogeneous dispersion of hexagonal boron nitride (h-BN) in epoxy matrix was achieved with a water-soluble carboxylated aniline trimer derivative (CAT{sup −}) as a dispersant, which was attributed to the strong π-π interaction between h-BN and CAT{sup −}, as proved by Raman and UV–vis spectra. Transmission electron microscopy (TEM) analysis confirmed a random dispersion of h-BN nanosheets in the waterborne epoxy coatings. The deterioration process of water-borne epoxy coating with and without h-BN nanosheets during the long-term immersion in 3.5 wt% NaCl solution was investigated by electrochemical measurements and water absorption test. Results implied that the introduction of well dispersed h-BN nanosheets into waterborne epoxy system remarkably improved the corrosion protection performance to substrate. Moreover, 1 wt% BN/EP composite coated substrate exhibited higher impedance modulus (1.3 × 10{sup 6} Ω cm{sup 2}) and lower water absorption (4%) than those of pure waterborne epoxy coating coated electrode after long-term immersion in 3.5 wt% NaCl solution, demonstrating its superior anticorrosive performance. This enhanced anticorrosive performance was mainly ascribed to the improved water barrier property of epoxy coating via incorporating homogeneously dispersed h-BN nanosheets.

PWSCC Preventive Maintenance Activities for Alloy 600 in Japanese PWR Plants

International Nuclear Information System (INIS)

Yamamoto, K.; Sugimoto, N.; Onishi, K.; Okimura, K.

2012-01-01

Because many nuclear plants have been in operation for ages, the importance of preventive maintenance technologies is getting higher. One conspicuous problem found in pressurized water reactor (PWR) plants is the primary water stress corrosion cracking (PWSCC) observed in Alloy 600 (a kind of high nickel based alloy) parts. Alloy 600 was used for butt welds between low alloy steel and stainless steel of nozzles of Reactor Vessel (RV), Steam Generator (SG), and Pressurizer (Pz). As PWSCC occurred at these parts may cause Loss of Coolant Accident (LOCA), preventive maintenance is necessary. PWSCC is considered to be caused by a mixture of three elements: high residual tensile stress on surface, material (Alloy 600) and environment. PWSCC can be prevented by improving one of the elements. MHI has been developing stress improvement methods, for example, Water Jet Peening (WJP), Shot Peening by Ultrasonic vibration (USP), and Laser Stress Improvement Process (L-SIP). According to the situation, appropriate method is applied for each part. WJP has been applied for RV nozzles of a lot of plants in Japan. However PWSCC was observed in RV nozzles during the inspection before WJP in recent years, MHI developed the Advanced INLAY system to improve the material from Alloy 600 to Alloy 690. Alloy 600 on the inner surface of the nozzles is removed and welding with Alloy 690 is performed. In addition, heat treatments for the nozzles are difficult for its structural situation, so ambient temperature temper bead welding technique for RV nozzles was developed to make the heat treatments unnecessary. This paper describes countermeasures against PWSCC and introduces the maintenance activities performed in Japan. (author)

Near-field heat transfer between graphene/hBN multilayers

Science.gov (United States)

Zhao, Bo; Guizal, Brahim; Zhang, Zhuomin M.; Fan, Shanhui; Antezza, Mauro

2017-06-01

We study the radiative heat transfer between multilayer structures made by a periodic repetition of a graphene sheet and a hexagonal boron nitride (hBN) slab. Surface plasmons in a monolayer graphene can couple with hyperbolic phonon polaritons in a single hBN film to form hybrid polaritons that can assist photon tunneling. For periodic multilayer graphene/hBN structures, the stacked metallic/dielectric array can give rise to a further effective hyperbolic behavior, in addition to the intrinsic natural hyperbolic behavior of hBN. The effective hyperbolicity can enable more hyperbolic polaritons that enhance the photon tunneling and hence the near-field heat transfer. However, the hybrid polaritons on the surface, i.e., surface plasmon-phonon polaritons, dominate the near-field heat transfer between multilayer structures when the topmost layer is graphene. The effective hyperbolic regions can be well predicted by the effective medium theory (EMT), thought EMT fails to capture the hybrid surface polaritons and results in a heat transfer rate much lower compared to the exact calculation. The chemical potential of the graphene sheets can be tuned through electrical gating and results in an additional modulation of the heat transfer. We found that the near-field heat transfer between multilayer structures does not increase monotonously with the number of layers in the stack, which provides a way to control the heat transfer rate by the number of graphene layers in the multilayer structure. The results may benefit the applications of near-field energy harvesting and radiative cooling based on hybrid polaritons in two-dimensional materials.

Thermal Conductivity of Epoxy Resin Composites Filled with Combustion Synthesized h-BN Particles

Directory of Open Access Journals (Sweden)

Shyan-Lung Chung

2016-05-01

Full Text Available The thermal conductivity of epoxy resin composites filled with combustion-synthesized hexagonal boron nitride (h-BN particles was investigated. The mixing of the composite constituents was carried out by either a dry method (involving no use of solvent for low filler loadings or a solvent method (using acetone as solvent for higher filler loadings. It was found that surface treatment of the h-BN particles using the silane 3-glycidoxypropyltrimethoxysilane (GPTMS increases the thermal conductivity of the resultant composites in a lesser amount compared to the values reported by other studies. This was explained by the fact that the combustion synthesized h-BN particles contain less –OH or active sites on the surface, thus adsorbing less amounts of GPTMS. However, the thermal conductivity of the composites filled with the combustion synthesized h-BN was found to be comparable to that with commercially available h-BN reported in other studies. The thermal conductivity of the composites was found to be higher when larger h-BN particles were used. The thermal conductivity was also found to increase with increasing filler content to a maximum and then begin to decrease with further increases in this content. In addition to the effect of higher porosity at higher filler contents, more horizontally oriented h-BN particles formed at higher filler loadings (perhaps due to pressing during formation of the composites were suggested to be a factor causing this decrease of the thermal conductivity. The measured thermal conductivities were compared to theoretical predictions based on the Nielsen and Lewis theory. The theoretical predictions were found to be lower than the experimental values at low filler contents (< 60 vol % and became increasing higher than the experimental values at high filler contents (> 60 vol %.

Theoretical predictions for hexagonal BN based nanomaterials as electrocatalysts for the oxygen reduction reaction.

Science.gov (United States)

Lyalin, Andrey; Nakayama, Akira; Uosaki, Kohei; Taketsugu, Tetsuya

2013-02-28

The catalytic activity for the oxygen reduction reaction (ORR) of both the pristine and defect-possessing hexagonal boron nitride (h-BN) monolayer and H-terminated nanoribbon have been studied theoretically using density functional theory. It is demonstrated that an inert h-BN monolayer can be functionalized and become catalytically active by nitrogen doping. It is shown that the energetics of adsorption of O(2), O, OH, OOH, and H(2)O on N atom impurities in the h-BN monolayer (N(B)@h-BN) is quite similar to that known for a Pt(111) surface. The specific mechanism of destructive and cooperative adsorption of ORR intermediates on the surface point defects is discussed. It is demonstrated that accounting for entropy and zero-point energy (ZPE) corrections results in destabilization of the ORR intermediates adsorbed on N(B)@h-BN, while solvent effects lead to their stabilization. Therefore, entropy, ZPE and solvent effects partly cancel each other and have to be taken into account simultaneously. Analysis of the free energy changes along the ORR pathway allows us to suggest that a N-doped h-BN monolayer can demonstrate catalytic properties for the ORR under the condition that electron transport to the catalytically active center is provided.

A novel combinatorial approach for the realization of advanced cBN composite coating

International Nuclear Information System (INIS)

Russell, W.C.; Yedave, S.N.; Sundaram, N.; Brown, W.D.; Malshe, A.P.

2001-01-01

The paper reports a novel coating process for the synthesis of hard material composite coatings. It consists of electrostatic spray coating (ESC) of powder particles (of micron-nanometer size) followed by chemical vapor infiltration (CVI) of a suitable binder phase. This novel approach enables fabrication of unique compositions such as cubic boron nitride (cBN) and titanium nitride (TiN) in a coating form. Recently, we have demonstrated the success of this technology by first coating a uniform over-layer (in excess of ∼ 10 μm) of cBN particles an carbide cutting tool inserts using ESC, followed by infiltration of particulate cBN matrix with TiN from its vapor phase using CVI to synthesize cBN-TiN a composite coating. The composite has shown excellent cBN-to-TiN and composite coating-to-carbide substrate adhesion. One of the main emphases of the paper is to discuss optimization and scale up of the ESC technology to achieve the desired microstructure and tailor the thickness across the cutting tool for better performance. Further, the cutting tools have been successfully tested for advanced machining applications. (author)

Calibration of thermal neutron detection compound BN-1 and CR-39 in the exposure room of Triga Reactor in Ljubljana

International Nuclear Information System (INIS)

Kristof, E.; Ilic, R.; Skvarc, J.; Dijanosic, R.

1994-01-01

Description of determination of thermal neutron fluences in the range from 1.E+02 to 1.E+12 cm -2 for calibration of the neutron sensitive compound consisting of the neutron converter BN-1 and charged particle detector CR-39 is given. The method employs two proportional BF3 detectors supplemented by a Ge(Li) gamma spectometer utilizing gold foils. The results of the measurements are also presented. (author)

Mechanical characterization of Si-C(O) fiber/SiC (CVI) matrix composites with a BN-interphase

International Nuclear Information System (INIS)

Prouhet, S.; Camus, G.; Labrugere, C.; Guette, A.; Martin, E.

1994-01-01

The mechanical behavior of three CVI-processed 2D woven SiC/BN/SiC composite materials with different initial BN interphase thicknesses has been investigated by means of tensile and impact tests. The results have established the efficiency of a BN interphase in promoting a nonlinear/noncatastrophic tensile behavior and high impact resistance. The effect of the initial BN interphase thickness on the resulting mechanical behavior has also been demonstrated. AES and TEM has revealed the presence of a SiO 2 /C double layer at the BN/fiber interface, which might result from a decomposition undergone by the Si-C(O) Nicalon fiber during processing. It has been suggested that the influence of the initial BN interphase thickness on the mechanical properties of the composites results from both changes occurring in the composition and morphology of the interfacial zones and modifications of the interfacial forces due to accommodation of the radial residual clamping stress

Monte Carlo Modeling Electronuclear Processes in Cascade Subcritical Reactor

CERN Document Server

Bznuni, S A; Zhamkochyan, V M; Polyanskii, A A; Sosnin, A N; Khudaverdian, A G

2000-01-01

Accelerator driven subcritical cascade reactor composed of the main thermal neutron reactor constructed analogous to the core of the VVER-1000 reactor and a booster-reactor, which is constructed similar to the core of the BN-350 fast breeder reactor, is taken as a model example. It is shown by means of Monte Carlo calculations that such system is a safe energy source (k_{eff}=0.94-0.98) and it is capable of transmuting produced radioactive wastes (neutron flux density in the thermal zone is PHI^{max} (r,z)=10^{14} n/(cm^{-2} s^{-1}), neutron flux in the fast zone is respectively equal PHI^{max} (r,z)=2.25 cdot 10^{15} n/(cm^{-2} s^{-1}) if the beam current of the proton accelerator is k_{eff}=0.98 and I=5.3 mA). Suggested configuration of the "cascade" reactor system essentially reduces the requirements on the proton accelerator current.

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

Energy Technology Data Exchange (ETDEWEB)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-02-26

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment

TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1996-01-01

An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

Concept of the new generation high safety liquid metal reactor (LMFR)

International Nuclear Information System (INIS)

Alekseev, P.N.; Zverkov, Y.A.; Morozov, A.G.; Orlov, V.V.; Ponomarev-Stepnoi, N.N.; Proshkin, A.A.; Slesarev, I.S.; Subbotin, S.A.

1988-01-01

The comparative analysis of the inner stability of the liquid metal reactors to severe accidents was made using the asymptotic reactivity balance. The group of the BN-reactors, Superphenix, IFR, LMFR were considered. This paper lists the characteristics of the reactors, used in the self-protectiveness analysis. The authors present the maximum coolant temperatures in post-accident asymptotic state for IFRs as on of the possible designs of a high safety fast reactor with metal fuel, U-Pu-Zr and LMFR. As is known, these values are very important for assessment of the ATWS accidence consequences. The authors consider the following situations and their combinations: loss of reactor coolant flow-LOFWS, loss of heat sink-LOHSWS, uncontrolled reactor sodium overcooling (down to the freezing point)-OVCWS, uncontrolled excess reactivity insertion-TOPWS. The calculation results demonstrate a high stability of the IFR and LMFR reactors to the most severe accidence sequences

Unraveling the structure of the h-BN/Rh(111) nanomesh with ab initio calculations

International Nuclear Information System (INIS)

Laskowski, R; Blaha, P

2008-01-01

The properties of a single layer of h-BN on top of a Rh(111) surface are discussed in terms of an ab initio generated force field approach as well as by direct ab initio density-functional theory (DFT) calculations. A single-layer model for the h-BN/Rh(111) nanomesh, in contrast to a previously considered (incomplete) double-layer model of h-BN, can explain the experimental data. The main focus of this work is to compare a force field approach described earlier in (Laskowski et al 2007 Phys. Rev. Lett. 98 106802) with direct ab initio calculations. The calculated geometry of the h-BN layer is very similar to the structure predicted by the force field approach. The ab initio calculated density of states projected on N-p x,y of BN corresponding to 'low' and 'high' regions with respect to the Rh surface shows a 1 eV splitting and thus explains the observed σ-band splitting. Moreover, we find good agreement between calculated and experimental scanning tunneling microscope (STM) images of this system

Electron Excess Doping and Effective Schottky Barrier Reduction on the MoS2/h-BN Heterostructure.

Science.gov (United States)

Joo, Min-Kyu; Moon, Byoung Hee; Ji, Hyunjin; Han, Gang Hee; Kim, Hyun; Lee, Gwanmu; Lim, Seong Chu; Suh, Dongseok; Lee, Young Hee

2016-10-12

Layered hexagonal boron nitride (h-BN) thin film is a dielectric that surpasses carrier mobility by reducing charge scattering with silicon oxide in diverse electronics formed with graphene and transition metal dichalcogenides. However, the h-BN effect on electron doping concentration and Schottky barrier is little known. Here, we report that use of h-BN thin film as a substrate for monolayer MoS 2 can induce ∼6.5 × 10 11 cm -2 electron doping at room temperature which was determined using theoretical flat band model and interface trap density. The saturated excess electron concentration of MoS 2 on h-BN was found to be ∼5 × 10 13 cm -2 at high temperature and was significantly reduced at low temperature. Further, the inserted h-BN enables us to reduce the Coulombic charge scattering in MoS 2 /h-BN and lower the effective Schottky barrier height by a factor of 3, which gives rise to four times enhanced the field-effect carrier mobility and an emergence of metal-insulator transition at a much lower charge density of ∼1.0 × 10 12 cm -2 (T = 25 K). The reduced effective Schottky barrier height in MoS 2 /h-BN is attributed to the decreased effective work function of MoS 2 arisen from h-BN induced n-doping and the reduced effective metal work function due to dipole moments originated from fixed charges in SiO 2 .

Correlation of Yield Stress And Microhardness in 08cr16ni11mo3 Irradiated To High Dose In The Bn-350 Fast Reactor

International Nuclear Information System (INIS)

Maksimkin, O.P.; Gusev, M.N.; Tivanova, O.S.; Silnaygina, N.S.; Garner, Francis A.

2006-01-01

The relationship between values of the microhardness and the engineering yield stress in steel 08Cr16Ni11Mo3 (Russian analog of AISI 316) heavily irradiated in the BN-350 reactor has been experimentally derived. It agrees very well with the previously published correlation developed by Toloczko for unirradiated 316 in a variety of cold-work conditions. Even more importantly, when the correlation is derived in the K δ format where the correlation involves changes in the two properties, we find excellent agreement with a universal K δ correlation developed by Busby and coworkers. With this K δ correlation, one can predict the value of yield stress in irradiated material based on measured values of microhardness. The technique is particularly suitable when the material of interest is in an inconvenient location or configuration, or when significant gradients in mechanical properties are anticipated over small dimensions. This approach makes it possible to reduce the labor input and risk when conducting such work. It appears that the derived correlation is equally applicable to both Russian and Western austenitic steel, and also in both irradiated and unirradiated conditions. Additionally, this report points out that microhardness measurements must take into account that high temperature sodium exposure alters the metal surface to produce ferrite, and therefore the altered layers should be removed prior to testing

Cleaning of the equipment of residual sodium by means of water-vacuum technology

International Nuclear Information System (INIS)

Klykov, B.P.; Lednev, A.I.

1997-01-01

Results of investigation into a problem of equipment decontamination from sodium, that have been conducted in OKBM since 1960 are given. The investigations performed have shown that a water-vacuum washing process is the most optimal method for equipment decontamination from sodium residues. The essence of the method is in conduction of sodium-water reaction under reduced pressure in a leak-tight tank. Boundary conditions are selected experimentally which not allow sodium to be melted during the process, that gives possibility to control the sodium-water reaction. Continuous removal of H 2 and reaction products creates safe conditions for the process conduction. More that 20-year period of operation of a stationary water-vacuum facility and washing the electromagnetic pump for BN-350 fast nuclear reactor directly at is test rig are the best proofs of the proposed method. This method is well suitable for washing the equipment contaminated by radioactive sodium, because by-products of the process are simply utilized. The method is used in a number of Russian enterprises, and recommended for implementation at BN-350 and BN-600 reactor plants. (author)

Elastic and Plastic Behavior of an Ultrafine-Grained Mg Reinforced with BN Nanoparticles

Science.gov (United States)

Trojanová, Zuzanka; Dash, Khushbu; Máthis, Kristián; Lukáč, Pavel; Kasakewitsch, Alla

2018-04-01

Pure microcrystalline magnesium (µMg) was reinforced with hexagonal boron nitride (hBN) nanoparticles and was fabricated by powder metallurgy process followed by hot extrusion. For comparison pure magnesium powder was consolidated by hot extrusion too. Both materials exhibited a significant fiber texture. Mg-hBN nanocomposites (nc) and pure Mg specimens were deformed between room temperature and 300 °C under tension and compression mode. The yield strength and ultimate tensile and compression strength as well as characteristic stresses were evaluated and reported. The tensile and compressive strengths of Mg-hBN nc are quiet superior in values compared to monolithic counterpart as well as Mg alloys. The compressive yield strength of µMg was recorded as 90 MPa, whereas the Mg-hBN nancomposite shows 125 MPa at 200 °C. The tensile yield strength of µMg was computed as 67 MPa which is quite lower as compared to Mg-hBN nanocomposite's value which was recorded as 157 MPa at 200 °C. Under tensile stress the true stress-strain curves are flat in nature, whereas the stress-strain curves observed in compression at temperatures up to 100 °C exhibited small local maxima at the onset of deformation followed by a significant work hardening.

Influence of microstructure in corrosion behavior of an Inconel 600 commercial alloy in 0.1 M sodium thiosulfate solution

International Nuclear Information System (INIS)

Granados, J.; Rodriguez, F.J.; Arganis, C.

1999-01-01

The Inconel 600 is used in diverse components of BWR and PWR type reactors, where diverse cases of intergranular stress corrosion have been presented. It has been reported susceptibility to the corrosion of this alloy, in presence of thiosulfates, which come from the degradation of the ion exchange resins of water treatments that use the reactors. The objective of this work is to study the influence of metallurgical condition in the corrosion velocity of Inconel 600 commercial alloy, in a 0.1 M thiosulfates solution. (Author)

111In-BnDTPA-F3: an Auger electron-emitting radiotherapeutic agent that targets nucleolin.

Science.gov (United States)

Cornelissen, Bart; Waller, Andrew; Target, Carol; Kersemans, Veerle; Smart, Sean; Vallis, Katherine A

2012-02-20

The F3 peptide (KDEPQRRSARLSAKPAPPKPEPKPKKAPAKK), a fragment of the human high mobility group protein 2, binds nucleolin. Nucleolin is expressed in the nuclei of normal cells but is also expressed on the membrane of some cancer cells. The goal was to investigate the use of 111In-labeled F3 peptide for Auger electron-targeted radiotherapy. F3 was labeled with fluorescein isothiocyanate (FITC) for confocal microscopy and conjugated to p-SCN-benzyl-diethylenetriaminepentaacetic acid (BnDTPA) for labeling with 111In to form 111In-BnDTPA-F3. MDA-MB-231-H2N (231-H2N) human breast cancer cells were exposed to 111In-BnDTPA-F3 and used in cell fractionation, γH2AX immunostaining (a marker of DNA double-strand breaks), and clonogenic assays. In vivo, biodistribution studies of 111In-BnDTPA-F3 were performed in 231-H2N xenograft-bearing mice. In tumor growth delay studies, 111In-BnDTPA-F3 (3 μg, 6 MBq/μg) was administered intravenously to 231-H2N xenograft-bearing mice once weekly for 3 weeks. Membrane-binding of FITC-F3 was observed in 231-H2N cells, and there was co-localization of FITC-F3 with nucleolin in the nuclei. After exposure of 231-H2N cells to 111In-BnDTPA-F3 for 2 h, 1.7% of 111In added to the medium was membrane-bound. Of the bound 111In, 15% was internalized, and of this, 37% was localized in the nucleus. Exposure of 231-H2N cells to 111In-BnDTPA-F3 (1 μM, 6 MBq/μg) resulted in a dose-dependent increase in γH2AX foci and in a significant reduction of clonogenic survival compared to untreated cells or cells exposed to unlabeled BnDTPA-F3 (46 ± 4.1%, 100 ± 1.8%, and 132 ± 7.7%, respectively). In vivo, tumor uptake of 111In-BnDTPA-F3 (3 μg, 6 MBq/μg) at 3-h post-injection was 1% of the injected dose per gram (%ID/g), and muscle uptake was 0.5%ID/g. In tumor growth delay studies, tumor growth rate was reduced 19-fold compared to untreated or unlabeled BnDTPA-F3-treated mice (p = 0.023). 111In-BnDTPA-F3 is internalized into 231-H2N cells and translocates

Enhanced Thermal Conductivity of Polyimide Composites Filled with Modified h-BN and Nanodiamond Hybrid Filler.

Science.gov (United States)

Yang, Xi; Yu, Xiaoyan; Naito, Kimiyoshi; Ding, Huili; Qu, Xiongwei; Zhang, Qingxin

2018-05-01

A new thermally conductive and electrically insulative polyimide were prepared by filling different amounts of hexagonal boron nitride (h-BN) particles, and the thermal conductivity of Polyimide (PI) composites were improved with the increasing h-BN content. Based on this, two methods were applied to improve thermal conductivity furtherly at limited filler loading in this paper. One is modifying the h-BN to improve interface interaction, another is fabricating a nano-micro hybrid filler with 2-D h-BN and 0-D nano-scale nanodiamond (ND) to build more effective conductive network. Both surface modification and hybrid system have a positive effect on thermal conductivity. The composites introducing 40 wt% hybrid filler (the weight ratio of ND/modified BN was 1/10) showed the highest thermal conductivity, being up to 0.98 W/(m K) (5.2 times that of PI). In addition, the composites exhibits excellent electrical insulation, thermal stability properties etc.

DFT simulation on H2 adsorption over Ni-decorated defective h-BN nanosheets

Science.gov (United States)

Zhou, Xuan; Chu, Wei; Zhou, Yanan; Sun, Wenjing; Xue, Ying

2018-05-01

Nickel doped defective h-BN nanosheets and their potential application on hydrogen storage were explored by density functional theory (DFT) calculation. Three types of defective h-BN (SW defect, VB and VN substrates) were modeled. In comparison with the SW defect, the B or N vacancy can improve the interaction between Ni atom and h-BN nanosheet strikingly. Furthermore, the Ni-doped SW defect sheet shows chemisorption on H2 molecules, and the Hsbnd H bond is partially dissociated. While on the VB sheet, Ni adatom interacts with H2 in the range of physisorption. However, the Ni-functionalized VN sheet exhibits a desirable adsorption on H2, and the corresponding energy varies from -0.40 to -0.51 eV, which is favorable for H2 adsorption and release at ambient conditions. As a result, the VN substrate is expected to a desirable support for H2 storage. Our work provides an insight into H2 storage on Ni-functionalized defective h-BN monolayer.

The modulation of Schottky barriers of metal-MoS2 contacts via BN-MoS2 heterostructures.

Science.gov (United States)

Su, Jie; Feng, Liping; Zhang, Yan; Liu, Zhengtang

2016-06-22

Using first-principles calculations within density functional theory, we systematically studied the effect of BN-MoS2 heterostructure on the Schottky barriers of metal-MoS2 contacts. Two types of FETs are designed according to the area of the BN-MoS2 heterostructure. Results show that the vertical and lateral Schottky barriers in all the studied contacts, irrespective of the work function of the metal, are significantly reduced or even vanish when the BN-MoS2 heterostructure substitutes the monolayer MoS2. Only the n-type lateral Schottky barrier of Au/BN-MoS2 contact relates to the area of the BN-MoS2 heterostructure. Notably, the Pt-MoS2 contact with n-type character is transformed into a p-type contact upon substituting the monolayer MoS2 by a BN-MoS2 heterostructure. These changes of the contact natures are ascribed to the variation of Fermi level pinning, work function and charge distribution. Analysis demonstrates that the Fermi level pinning effects are significantly weakened for metal/BN-MoS2 contacts because no gap states dominated by MoS2 are formed, in contrast to those of metal-MoS2 contacts. Although additional BN layers reduce the interlayer interaction and the work function of the metal, the Schottky barriers of metal/BN-MoS2 contacts still do not obey the Schottky-Mott rule. Moreover, different from metal-MoS2 contacts, the charges transfer from electrodes to the monolayer MoS2, resulting in an increment of the work function of these metals in metal/BN-MoS2 contacts. These findings may prove to be instrumental in the future design of new MoS2-based FETs with ohmic contact or p-type character.

Cubic boron nitride (cBN) - A new material for advanced optoelectronic devices. Properties and perspectives

International Nuclear Information System (INIS)

Nistor, S.V.; Nistor, L.C.; Dinca, G.

2001-01-01

Cubic boron nitride (cBN) exhibits, besides exceptional thermal and mechanical properties similar to diamond, an excellent ability to be n or p doped, which makes it a strong candidate for advanced, high - temperature optical and microelectronic devices. Despite its outstanding characteristics, there are quite a few reports concerning the physical properties of cBN. This is partly due to the absence of natural cBN gems and the extreme difficulties in producing enough large (mm 3 sized) single crystals, or single phase thin films, for physical characterization. The state of the art knowledge concerning the basic properties of crystalline cBN, as well as our recent results of microstructure and defect properties studies will be presented. (authors)

Design studies on staffing requirements for the new generation nuclear power units of WWER-640 and BN-800 reactor types

International Nuclear Information System (INIS)

Solovyov, D.F.

2001-01-01

The paper outlines the main staffing requirements for the new generation power units with nuclear reactors. These requirements were developed taking into account IAEA recommendations. NPP staffing structure is described, including the main and auxiliary personnel. The main principles of personnel number determination are given. Special attention is taken to the issues of personnel skill and training, including both theoretical education and practical work on the power units in operation. The use of simulators, system of knowledge control and structure of training are considered. ''Shopless'' staffing structure approach is proposed for the NPP, assuming that the main scope of repair work is performed by the central repair organization, thus increasing the quality of repair and decreasing the number of personnel on the plant. Data are given on the personnel number for the WWER-640 and the BN-800 reactor designs. Specialists of the ''ATOMENERGOPROJECT'' Institute started their work on staffing on the early development stage of the basic design of WWER-640 reactor power unit which is the forerunner of the new generation reactors. This work was based on the approaches taken by the chief engineers of NPPs in operation during their meeting held in 1989 in Kalinin NPP. At this meeting definite decision was taken on changing over to involving manufacturer in the repair work of NPP components using manufacturer's technology. In 1992 the meeting of representatives of suppliers of the main components was held where representatives of ''ATOMENERGOREMONT'' and ''LENENERGOREMONT'' were present. The suppliers agreed on carrying out repair works on the components they produced. For this purpose special departments were set up having some experience. This repair work is already carried out by ''ATOMENERGOREMONT'' on some nuclear power plants. ''LENENERGOREMONT'' has gained considerable experience in this kind of repair work on the turbines of LO-1 and LO-2 NPP in Finland. Within the

The status of safeguarding 600 MW(e) CANDU reactors

International Nuclear Information System (INIS)

Von Baeckmann, A.; Rundquist, D.E.; Pushkarjov, V.; Smith, R.M.; Zarecki, C.W.

1982-09-01

There has been extensive work in the development of CANDU safeguards since the last International Conference on Nuclear Power, and this has resulted in the development of improved equipment for the safeguards system now being installed in the 600 MW(e) CANDU generating stations. The overall system is designed to improve on the existing IAEA safeguards and to provide adequate coverage for each plausible nuclear material diversion route. There is sufficient sensitivity and redundancy to enable the timely detection of the possible diversion of significant quantities of nuclear material

Estimating residual life of alloy 600 RPV penetrations

International Nuclear Information System (INIS)

Hunt, E.S.; White, G.A.; Pathania, R.; Arey, M.L.; Whitaker, D.E.

1996-01-01

Primary water stress corrosion cracking (PWSCC) of Alloy 600 penetrations PWR in reactor pressure vessel (RPV) heads has become a significant economic concern worldwide. PWSCC of these penetrations has led to extended maintenance outages, expensive inspections and repairs, and in some cases, replacement of the entire vessel head. This paper describes methodology developed to predict the remaining life of Alloy 600 penetrations in reactor vessel heads. Predictions of remaining life are an important input to planning models used by utilities to select a strategy for responding to the PWSCC issue at the lowest life cycle cost with an acceptably low risk of leakage. The remaining life of RPV penetrations is determined using the results of inspections of penetrations and statistical methods to predict future degradation. The analysis takes into account the effects of material properties, welding residual stresses, and operating temperature on PWSCC initiation and growth. The probability of developing cracks of various depths is assessed using Monte Carlo methods which provide for uncertainties in the input assumptions. For plants which have not yet performed inspections, remaining life predictions are based on inspection results from similar plants which have performed inspections with corrections made for known differences in design details, material properties and operating conditions

Inherently safe reactors

International Nuclear Information System (INIS)

Maartensson, Anders

1992-01-01

A rethinking of nuclear reactor safety has created proposals for new designs based on inherent and passive safety principles. Diverging interpretations of these concepts can be found. This article reviews the key features of proposed advanced power reactors. An evaluation is made of the degree of inherent safety for four different designs: the AP-600, the PIUS, the MHTGR and the PRISM. The inherent hazards of today's most common reactor principles are used as reference for the evaluation. It is concluded that claims for the new designs being inherently, naturally or passively safe are not substantiated by experience. (author)

Safety features and research needs of westinghouse advanced reactors

International Nuclear Information System (INIS)

Carelli, M.D.; Winters, J.W.; Cummins, W.E.; Bruschi, H.J.

2002-01-01

The three Westinghouse advanced reactors - AP600, AP1000 and IRIS - are at different levels of readiness. AP600 has received a Design Certification, its larger size version AP1000 is currently in the design certification process and IRIS has just completed its conceptual design and will initiate soon a licensing pre-application. The safety features of the passive designs AP600/AP1000 are presented, followed by the features of the more revolutionary IRIS, a small size modular integral reactor. A discussion of the IRIS safety by design approach is given. The AP600/AP1000 design certification is backed by completed testing and development which is summarized, together with a research program currently in progress which will extend AP600 severe accident test data to AP1000 conditions. While IRIS will of course rely on applicable AP600/1000 data, a very extensive testing campaign is being planned to address all the unique aspects of its design. Finally, IRIS plans to use a risk-informed approach in its licensing process. (authors)

Methodologies to assess PWSCC susceptibility of primary component Alloy 600 locations in pressurized water reactors

International Nuclear Information System (INIS)

Rao, G.V.

1993-01-01

Methodologies to assess susceptibility to Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600 component locations in the Primary System of Pressurized Water Reactors are presented. The assessment methodologies are presented. The assessment methodologies are based on Relative Susceptibility Index (RSI) and Cumulative Susceptibility Index (CSI) models utilizing key contributing parameters such as service and residual stresses, yield strength, service temperature, material condition and microstructure, and the accumulated service time. To aid in the development of future inspection plans, a method of ranking of the assessed susceptibilities by 'bench marking' with respect to the susceptibility of a reference location of known PWSCC history of a reference location of known PWSCC history is presented. Means of utilizing the susceptibility ranking results in developing a prioritized inspection plan are discussed. A follow-up investigative plan to the initial inspection is proposed, which includes identification of critical sampling locations, sample extraction, sample investigations and testing to ensure that the potentially highest susceptibility locations are free from near term PWSCC and, further, to provide a basis for established schedules for future inspections. Finally, parametric considerations of the contributing factor are presented to help the utility choose suitable option to mitigate the PWSCC issue while minimizing the impact on continued service

Calculation of the octanol-water partition coefficient of armchair polyhex BN nanotubes

Science.gov (United States)

Mohammadinasab, E.; Pérez-Sánchez, H.; Goodarzi, M.

2017-12-01

A predictive model for determination partition coefficient (log P) of armchair polyhex BN nanotubes by using simple descriptors was built. The relationship between the octanol-water log P and quantum chemical descriptors, electric moments, and topological indices of some armchair polyhex BN nanotubes with various lengths and fixed circumference are represented. Based on density functional theory electric moments and physico-chemical properties of those nanotubes are calculated.

Applicability of PRISM PRA Methodology to the Level II Probabilistic Safety Analysis of KALIMER-600 (I) (Core Damage Event Tree Analysis Part)

International Nuclear Information System (INIS)

Park, S. Y.; Kim, T. W.; Ha, K. S.; Lee, B. Y.

2009-03-01

The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the PSA domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of KALIMER-600 reactor. An applicability of the PSA of the PRISM plant to the KALIMER-600 has been studied. The study is confined to a core damage event tree analysis which is a part of a level 2 PSA. Assuming that the accident types, which can be developed from level 1 PSA, are same as the PRISM PRA, core damage categories are defined and core damage event trees are developed for the KALIMER-600 reactor. Fission product release fractions of the core damage categories and branch probabilities of the core damage event trees are referred from the PRISM PRA temporarily. Plant specific data will be used during the detail analysis

Acoustically-driven surface and hyperbolic plasmon-phonon polaritons in graphene/h-BN heterostructures on piezoelectric substrates

Science.gov (United States)

Fandan, R.; Pedrós, J.; Schiefele, J.; Boscá, A.; Martínez, J.; Calle, F.

2018-05-01

Surface plasmon polaritons in graphene couple strongly to surface phonons in polar substrates leading to hybridized surface plasmon-phonon polaritons (SPPPs). We demonstrate that a surface acoustic wave (SAW) can be used to launch propagating SPPPs in graphene/h-BN heterostructures on a piezoelectric substrate like AlN, where the SAW-induced surface modulation acts as a dynamic diffraction grating. The efficiency of the light coupling is greatly enhanced by the introduction of the h-BN film as compared to the bare graphene/AlN system. The h-BN interlayer not only significantly changes the dispersion of the SPPPs but also enhances their lifetime. The strengthening of the SPPPs is shown to be related to both the higher carrier mobility induced in graphene and the coupling with h-BN and AlN surface phonons. In addition to surface phonons, hyperbolic phonons polaritons (HPPs) appear in the case of multilayer h-BN films leading to hybridized hyperbolic plasmon-phonon polaritons (HPPPs) that are also mediated by the SAW. These results pave the way for engineering SAW-based graphene/h-BN plasmonic devices and metamaterials covering the mid-IR to THz range.

Effect of substrate temperature and gas flow ratio on the nanocomposite TiAlBN coating

Energy Technology Data Exchange (ETDEWEB)

Rosli, Z. M., E-mail: azmr@utem.edu.my; Kwan, W. L., E-mail: kwailoon86@gmail.com; Juoi, J. M., E-mail: jariah@utem.edu.my [Faculty of Manufacturing Engineering, Universiti Teknikal Malaysia Melaka, Hang Tuah Jaya, 76100 Durian Tunggal, Melaka (Malaysia)

2016-07-19

Nanocomposite TiAlBN (nc-TiAlBN) coatings were successfully deposited via RF magnetron sputtering by varying the nitrogen-to-total gas flow ratio (R{sub N}), and substrate temperature (T{sub S}). All coatings were deposited on AISI 316 substrates using single Ti-Al-BN hot-pressed disc as a target. The grain size, phases, and chemical composition of the coatings were evaluated using glancing angle X-ray diffraction analysis (GAXRD) and X-ray photoelectron spectroscopy (XPS). Results showed that the grains size of the deposited nc-TiAlBN coatings were in the range of 3.5 to 5.7 nm and reached a nitride saturation state as early as 15 % R{sub N}. As the nitrogen concentration decreases, boron concentration increased from 9 at.% to 16.17 at.%. and thus, increase the TiB{sub 2} phase within the coatings. The T{sub S}, however, showed no significant effect either on the crystallographic structure, grain size, or in the chemical composition of the deposited nc-TiAlBN coating.

Effect of substrate temperature and gas flow ratio on the nanocomposite TiAlBN coating

International Nuclear Information System (INIS)

Rosli, Z. M.; Kwan, W. L.; Juoi, J. M.

2016-01-01

Nanocomposite TiAlBN (nc-TiAlBN) coatings were successfully deposited via RF magnetron sputtering by varying the nitrogen-to-total gas flow ratio (R_N), and substrate temperature (T_S). All coatings were deposited on AISI 316 substrates using single Ti-Al-BN hot-pressed disc as a target. The grain size, phases, and chemical composition of the coatings were evaluated using glancing angle X-ray diffraction analysis (GAXRD) and X-ray photoelectron spectroscopy (XPS). Results showed that the grains size of the deposited nc-TiAlBN coatings were in the range of 3.5 to 5.7 nm and reached a nitride saturation state as early as 15 % R_N. As the nitrogen concentration decreases, boron concentration increased from 9 at.% to 16.17 at.%. and thus, increase the TiB_2 phase within the coatings. The T_S, however, showed no significant effect either on the crystallographic structure, grain size, or in the chemical composition of the deposited nc-TiAlBN coating.

Structural, electronic, and magnetic properties of 3D metal trioxide and tetraoxide superhalogen cluster-doped monolayer BN

International Nuclear Information System (INIS)

Meng, Jingjing; Li, Dan; Niu, Yuan; Zhao, Hongmin; Liang, Chunjun; He, Zhiqun

2016-01-01

The structural, electronic, and magnetic properties of monolayer BN doped with 3D metal trioxide and tetraoxide superhalogen clusters are investigated using first-principle calculations. TMO_3_(_4_)-doped monolayer BN exhibits a low negative formation energy, whereas TM atoms embedded in monolayer BN show a high positive formation energy. TMO_3_(_4_) clusters are embedded more easily in monolayer BN than TM atoms. Compared with TMO_3-doped structures, TMO_4-doped structures have a higher structural stability because of their higher binding energies. Given their low negative formation energies, TMO_4-doped structures are more favored for specific applications than TMO_3-doped structures and TM atom-doped structures. Large magnetic moments per supercell and significant ferromagnetic couplings between a TM atom and neighboring B and N atoms on the BN layer were observed in all TMO_4-doped structures, except for TiO_4-doped structures. - Highlights: • TMO_3_(_4_) superhalogen clusters incorporated into monolayer BN were investigated. • TMO_3_(_4_) clusters are embedded more easily in monolayer BN than TM atoms. • TMO_4-doped structures are more favored for specific applications. • Large magnetic moments were observed in TMO_4-doped structures. • The band gap was sensitively dependent on the doped clusters.

Localized corrosion problems in water reactors

International Nuclear Information System (INIS)

Coriou, Henri.

1977-01-01

Main localized etching on the structure materials of water reactors are studied: stress corrosion on stainless steel 304 (B.W.R), stress corrosion, 'wall thinning' and denting of Inconel 600 vapor generator tubes (P.W.R.). Some mechanisms are examined and practical exemples in reactors are described. Various possible cures are presented [fr

Preliminary Assessment of PHTS Pump Piping Break Accident of DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Choi, Yongwon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

KINS is evaluating the applicability of TRACE code for safety analysis of SFR Since 2012. Based on the steady-state input deck for Demonstration Sodium Cooled Fast Reactor 600MW (DSFR-600) component-wise specific modeling is developed for DSFR-600. Preliminary analysis was performed with TRACE code for DSFR-600 PHTS pump piping break accident. The calculation result showed that the calculated safety parameters are conforms to the design criteria for DBA accidents. RHRS design of DSFR-600 and its performance during transient was also reviewed by sensitivity study on the effect of sodium condition to the transient decay heat removal capability of RHRS. Following insights are identified. These should be considered in improving the design also in licensing review of SFR safety analysis. The transient performance of RHRS might differ from the component's design capacity. RHRS's transient performance also should be included in the design documents and validated with reasonable test and/or analysis with consideration of the variation of coolant conditions during transient. The analytic model used for safety analysis should consider 3-D effect of vessel pool and its uncertainty with reasonable conservatism.

Oxidation of BN-coated SiC fibers in ceramic matrix composites

International Nuclear Information System (INIS)

Sheldon, B.W.; Sun, E.Y.

1996-01-01

Thermodynamic calculations were performed to analyze the simultaneous oxidation of BN and SiC. The results show that, with limited amounts of oxygen present, the formation of SiO 2 should occur prior to the formation of B 2 O 3 . This agrees with experimental observations of oxidation in glass-ceramic matrix composites with BN-coated SiC fibers, where a solid SiO 2 reaction product containing little or no boron has been observed. The thermodynamic calculations suggest that this will occur when the amount of oxygen available is restricted. One possible explanation for this behavior is that SiO 2 formation near the external surfaces of the composite closes off cracks or pores, such that vapor phase O 2 diffusion into the composite occurs only for a limited time. This indicates that BN-coated SiC fibers will not always oxidize to form significant amounts of a low-melting, borosilicate glass

Commercializing the next generation: the AP600 advanced simplified nuclear power plant

International Nuclear Information System (INIS)

Bruschi, H.J.

1994-01-01

Today, government and industry are working together on advanced nuclear power plant designs that take advantage of valuable lessons learned from the experience to date and promise to reconcile the demands of economic expansion with the laws of environmental protection. In the U.S., the Department of Energy (DOE) and the Electric Power Research Institute (EPRI) initiated a design certification program in 1989 to develop and commercialize advanced light water reactors (ALWRs) for the next round of power plant construction. Advanced, simplified technology is one approach under development to end the industry's search for a simpler, more forgiving, and less costly reactor. As part of this program, Westinghouse is developing the AP600, a new standard 600 MWe advanced, simplified plant. The design strikes a balance between the use of proven technology and new approaches. The result is a greatly streamlined plant that can meet safety regulations and reliability requirements, be economically competitive, and promote broader public confidence in nuclear energy. 1 fig

Fabrication of BN/Al(-Mg) metal matrix composite (MMC) by pressureless infiltration technique

Energy Technology Data Exchange (ETDEWEB)

Jung, W.G.; Kwon, H. [School of Advanced Materials Eng., Kookmin Univ., Seoul (Korea)

2004-07-01

BN/Al(-Mg) metal matrix composite (MMC) was fabricated by the pressureless infiltration technique. The phase characterizations of the composites were analyzed using the SEM, TEM, EDS and EPMA on reaction products after the electrochemical dissolution of the matrix. It is confirmed that aluminum nitride (AlN) was formed by the reaction of Mg{sub 3}N{sub 2} and Al alloy melt. Plate type AlN and polyhedral type Mg(-Al) boride were formed by the reaction between Mg{sub 3}N{sub 2}, BN and molten Al in the composite. The reaction mechanism in the fabrication of BN/Al(-Mg) MMC was derived from the phase analysis results and the thermodynamic investigation. (orig.)

Investigation of structure and mechanical properties of plasma vapor deposited nanocomposite TiBN films

Science.gov (United States)

Han, Bin; Neena, D.; Wang, Zesong; Kondamareddy, K. k.; Li, Na; Zuo, Wenbin; Yan, Shaojian; Liu, Chuansheng; Fu, Dejun

2017-04-01

TiBN coatings have huge potential applications as they have excellent properties with increasing modern industrial requirements. Nanocomposite TiBN coatings were synthesized on cemented carbide, high speed steel and Si substrates by using cathodic arc plasma ion plating from pure TiB2 ceramic targets. The structure and mechanical properties of the TiBN coatings were significantly influenced by the nitrogen partial pressure. Rutherford backscattering spectrometry demonstrates that the nitrogen content of the coating varied from 2.8% to 34.5% and high-resolution electron microscopy images reveal that all coatings have the characteristic of nanocrystals embedded in an amorphous matrix. The root-mean-square roughness of the coatings increases from 3.73 to 14.64 nm and the coefficients of friction of the coatings at room temperature vary from 0.54 to 0.73 with increasing nitrogen partial pressure. The microhardness of the coating increases up to 35.7 GPa at 10 sccm N2 flow rate. The smallest wear rate is 2.65 × 10-15 m3 N-1 m-1 which indicates that TiBN coatings have excellent wear resistance. The adhesion test revealed that the TiBN coatings have good adhesion at low nitrogen partial pressure.

Velocity of crack growing of Inconel-600, sensitized, contaminated with sulphur in PWR type reactors

International Nuclear Information System (INIS)

Castano, M. L.; Blazquez, F.; Gomez Briceno, D.; Lagares, A.

1998-01-01

The origin of the vessel head penetration cracking of Jose Cabrera NPP has been attributed to an IGA/SCC process in a highly sensitized Alloy 600 assisted by sulphur species, as both acid sulphates and reduced species originated by the thermal breakdown of the cationic resins present in the primary coolant. The thermal degradation of the cationic resins leads sulphonic acid group scission and sulphates. Under the operating conditions the reduction of sulphates to sulphides is produced. The sulphides formed from the reduction of sulphate can precipitate with metallic cations and be incorporated into the oxide layers of the materials, preferably into nickel alloys. Others components at Jose Cabrera NPP are fabricated from sensitized alloy 600, as bottom vessel penetrations. In order to determine the influence of sulphur incorporated to the oxide layers of bottom vessel penetration alloy 600, an experimental work has been performed to obtained crack growth rate data under PWR primary conditions on sensitized alloy 600. (Author) 5 refs

A nano capacitor with graphene electrodes and Methane - (h-BN)insulator

OpenAIRE

Farrokh Roya Nikmaram

2016-01-01

Methan has a large potential to adsorb and diffuse among h-BN and graphene surfaces as the suitable dielectric. With this background the nanoscale dielectric capacitors have been widely studied due to their ability to store a high amount of energy. In this research, I have modeled one which is composed of two graphene layers including insulating medium of a h-BN layers which are filed out (Methane)n,m {n=m=7). It has been indicated thatthe Methane moleculeis the suitable gas for hetero-struct...

Plasma synthesis and HPHT consolidation of BN nanoparticles, nanospheres, and nanotubes to produce nanocrystalline cubic boron nitride

Science.gov (United States)

Stout, Christopher

Plasma methods offer a variety of advantages to nanomaterials synthesis. The process is robust, allowing varying particle sizes and phases to be generated simply by modifying key parameters. The work here demonstrates a novel approach to nanopowder synthesis using inductively-coupled plasma to decompose precursor, which are then quenched to produce a variety of boron nitride (BN)-phase nanoparticles, including cubic phase, along with short-range-order nanospheres (e.g., nano-onions) and BN nanotubes. Cubic BN (c-BN) powders can be generated through direct deposition onto a chilled substrate. The extremely-high pyrolysis temperatures afforded by the equilibrium plasma offer a unique particle growth environment, accommodating long deposition times while exposing resulting powders to temperatures in excess of 5000K without any additional particle nucleation and growth. Such conditions can yield short-range ordered amorphous BN structures in the form of 20nm diameter nanospheres. Finally, when introducing a rapid-quenching counter-flow gas against the plasma jet, high aspect ratio nanotubes are synthesized, which are collected on substrate situated radially. The benefits of these morphologies are also evident in high-pressure/high-temperature consolidation experiments, where nanoparticle phases can offer a favorable conversion route to super-hard c-BN while maintaining nanocrystallinity. Experiments using these morphologies are shown to begin to yield c-BN conversion at conditions as low as 2.0 GPa and 1500°C when using micron sized c-BN seeding to create localized regions of high pressures due to Hertzian forces acting on the nanoparticles.

Nanowires and nanotubes of BN, GaN and Si3N4

International Nuclear Information System (INIS)

Deepak, F.L.; Gundiah, G.; Govindaraj, A.; Rao, C.N.

2002-01-01

Simple methods of synthesizing nanotubes and nanowires of boron nitride, gallium nitride and silicon nitride have been investigated. The nanotubes and nanowires have been examined by electron microscopy and other techniques. In the case of BN, activated carbon or multi-walled carbon nanotubes (MWNTs) was heated with boric acid in the presence of NH 3 . With activated carbon, BN nanowires constitute the primary products, but good yields of BN nanotubes are obtained with MWNTs. Aligned BN nanotubes are obtained when aligned MWNTs are employed as the starting material suggesting templating role of carbon nanotubes. Single crystal gallium nitride nanowires have been obtained by heating carbon nanotubes coated with gallium acetylacetonate in NH 3 vapor at 910 o C. Single walled carbon nanotubes were used as templated to reduce the diameter of the GaN nanowires. The growth direction of the GaN nanowires is nearly perpendicular to the [100] planes and the nanowires exhibit satisfactory photoluminescence spectra. Si 3 N 4 nanowires have been synthesized by heating multi-walled carbon nanotubes with silica gel at 1360 o C in an atmosphere of NH 3 . Si 3 N 4 nanotubes are found occasionally when aligned multi-walled nanotubes are employed as templates. (author)

SPES-2, an experimental program to support the AP600 development

Energy Technology Data Exchange (ETDEWEB)

Tarantini, M. [ENEA, Nuclear Fission Branch, Bolonga (Italy); Medich, C. [SIET S.p.A. Piacenza (Italy)

1995-09-01

In support of the development of the AP600 reactor, ENEA, ENEL, ANSALDO and Westinghouse have signed a research agreement. In the framework of this agreement a complex Full Height Full Pressure (FHFP) integral system testing program has been planned on SPES-2 facility. The main purpose of this paper is to point out the status of the test program; describe the hot per-operational test performed and the complete test matrix, giving all the necessary references on the work already published. Two identical Small Break LOCA transients, performed with Pressurizer to Core Make-up Tank (PRZ-CMT) balance line (Test S00203) and without PRZ-CMT balance line (Test S00303) are then compared, to show how the SPES-2 facility can contribute in confirming the new AP600 reactor design choices and can give useful indications to designers. Although the detailed analysis of test data has not been completed, some consideration on the analytical tools utilized and on the SPES-2 capability to simulate the reference plant is then drawn.

Effects of hexagonal boron nitride and sintering temperature on mechanical and tribological properties of SS316L/h-BN composites

International Nuclear Information System (INIS)

Mahathanabodee, S.; Palathai, T.; Raadnui, S.; Tongsri, R.; Sombatsompop, N.

2013-01-01

Highlights: ► 20 vol% h-BN in stainless steel gave the lowest friction coefficient. ► Sintering temperature of 1200 °C was recommended for optimum friction coefficient. ► h-BN in stainless steel transformed to a boride liquid phase at 1250 °C. - Abstract: In this work, hexagonal boron nitride (h-BN)-embedded 316L stainless steel (SS316L/h-BN) composites were prepared using a conventional powder metallurgy process. In order to produce self-lubricating composites, various amounts of h-BN (10, 15 and 20 vol%) were incorporated. Effects of h-BN content and sintering temperature on the mechanical and tribological properties were of primary interest. The results suggested that an increase in h-BN content reduced the hardness of the composites, but that the hardness could be improved by increasing the sintering temperature. Addition of h-BN up to 20 vol% improved the friction coefficient of the composites. At a sintering temperature of 1250 °C, h-BN transformed into a boride liquid phase, which formed a eutectic during cooling and exhibited a deterioration effect on lubricating film formation of the h-BN, resulting in an increase in the friction coefficient of the composites. The specific wear rate was greatly reduced when the composites were sintered at 1200 °C. The lowest friction coefficient and specific wear rate in the composites could be found under the experimental conditions used in this work when using 20 vol% of h-BN at a sintering temperature of 1200 °C

Suppression of Lateral Diffusion and Surface Leakage Currents in nBn Photodetectors Using an Inverted Design

Science.gov (United States)

Du, X.; Savich, G. R.; Marozas, B. T.; Wicks, G. W.

2018-02-01

Surface leakage and lateral diffusion currents in InAs-based nBn photodetectors have been investigated. Devices fabricated using a shallow etch processing scheme that etches through the top contact and stops at the barrier exhibited large lateral diffusion current but undetectably low surface leakage. Such large lateral diffusion current significantly increased the dark current, especially in small devices, and causes pixel-to-pixel crosstalk in detector arrays. To eliminate the lateral diffusion current, two different approaches were examined. The conventional solution utilized a deep etch process, which etches through the top contact, barrier, and absorber. This deep etch processing scheme eliminated lateral diffusion, but introduced high surface current along the device mesa sidewalls, increasing the dark current. High device failure rate was also observed in deep-etched nBn structures. An alternative approach to limit lateral diffusion used an inverted nBn structure that has its absorber grown above the barrier. Like the shallow etch process on conventional nBn structures, the inverted nBn devices were fabricated with a processing scheme that only etches the top layer (the absorber, in this case) but avoids etching through the barrier. The results show that inverted nBn devices have the advantage of eliminating the lateral diffusion current without introducing elevated surface current.

Band gap tunning in BN-doped graphene systems with high carrier mobility

KAUST Repository

Kaloni, T. P.

2014-02-17

Using density functional theory, we present a comparative study of the electronic properties of BN-doped graphene monolayer, bilayer, trilayer, and multilayer systems. In addition, we address a superlattice of pristine and BN-doped graphene. Five doping levels between 12.5% and 75% are considered, for which we obtain band gaps from 0.02 eV to 2.43 eV. We demonstrate a low effective mass of the charge carriers.

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

International Nuclear Information System (INIS)

Bednyakov, Sergey

2000-12-01

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)

Vacuolar iron transporter BnMEB2 is involved in enhancing iron tolerance of Brassica napus

Directory of Open Access Journals (Sweden)

Wei Zhu

2016-09-01

Full Text Available Iron toxicity is a major nutrient disorder that severely affects crop development and yield. Vacuolar detoxification of metal stress is an important strategy for plants to survive and adapt to this adverse environment. Vacuolar iron transporter (VIT members are involved in this process and play essential roles in iron storage and transport. In this study, a rapeseed VIT gene BnMEB2 (BnaC07g30170D was identified. BnMEB2 is a homolog to Arabidopsis MEB2 (At5g24290 and acts as a detoxifier in vacuolar sequestration of divalent metal. Transient expression analysis revealed that BnMEB2 was localized to the vacuolar membrane. Q-PCR detection showed a high expression of BnMEB2 in mature (60-day-old leaves and could be obviously induced by exogenous iron stress in both roots and leaves. Over-expressed BnMEB2 in both Arabidopsis wild type and meb2 mutant seedlings resulted in greatly improved iron tolerability with no significant changes in the expression level of other vacuolar iron transporter genes. The mutant meb2 grew slowly and its root hair elongation was inhibited under high iron concentration condition while BnMEB2 over-expressed transgenic plants of the mutant restored the phenotypes with apparently higher iron storage in roots and dramatically increased iron content in the whole plant. Taken together, these results suggested that BnMEB2 was a VIT gene in rapeseed which was necessary for safe storage and vacuole detoxification function of excess iron to enhance the tolerance of iron toxicity. This research sheds light on a potentially new strategy for attenuating hazardous metal stress from environment and improving iron biofortification in Brassicaceae crops.

Exfoliated BN shell-based high-frequency magnetic core-shell materials.

Science.gov (United States)

Zhang, Wei; Patel, Ketan; Ren, Shenqiang

2017-09-14

The miniaturization of electric machines demands high frequency magnetic materials with large magnetic-flux density and low energy loss to achieve a decreased dimension of high rotational speed motors. Herein, we report a solution-processed high frequency magnetic composite (containing a nanometal FeCo core and a boron nitride (BN) shell) that simultaneously exhibits high electrical resistivity and magnetic permeability. The frequency dependent complex initial permeability and the mechanical robustness of nanocomposites are intensely dependent on the content of BN insulating phase. The results shown here suggest that insulating magnetic nanocomposites have potential for application in next-generation high-frequency electric machines with large electrical resistivity and permeability.

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

International Nuclear Information System (INIS)

Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

2006-01-01

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A and M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900-1000 deg. C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900-1000 deg. C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and

The BnALMT1 Protein That is an Aluminum-Activated Malate Transporter is Localized in the Plasma Membrane

OpenAIRE

Ligaba, Ayalew; Katsuhara, Maki; Sakamoto, Wataru; Matsumoto, Hideaki

2007-01-01

We have previously reported that Al-induces citrate and malate efflux from P-sufficient and P-deficient plants of rape (Brassica napus L.) and that P-deficiency alone could not induce this response. Further investigation showed that the transcript of two genes designated BnALMT1 and BnALMT2 is accumulated in roots by Al-treatment. Transgenic tobacco cells (Nicotiana tabacum) and Xenopus laevis oocytes expressing the BnALMT1 and BnALMT2 proteins released more malate than control cells in the p...

nBn Infrared Detector Containing Graded Absorption Layer

Science.gov (United States)

Gunapala, Sarath D.; Ting, David Z.; Hill, Cory J.; Bandara, Sumith V.

2009-01-01

It has been proposed to modify the basic structure of an nBn infrared photodetector so that a plain electron-donor- type (n-type) semiconductor contact layer would be replaced by a graded n-type III V alloy semiconductor layer (i.e., ternary or quarternary) with appropriate doping gradient. The abbreviation nBn refers to one aspect of the unmodified basic device structure: There is an electron-barrier ("B" ) layer between two n-type ("n" ) layers, as shown in the upper part of the figure. One of the n-type layers is the aforementioned photon-absorption layer; the other n-type layer, denoted the contact layer, collects the photocurrent. The basic unmodified device structure utilizes minority-charge-carrier conduction, such that, for reasons too complex to explain within the space available for this article, the dark current at a given temperature can be orders of magnitude lower (and, consequently, signal-to-noise ratios can be greater) than in infrared detectors of other types. Thus, to obtain a given level of performance, less cooling (and, consequently, less cooling equipment and less cooling power) is needed. [In principle, one could obtain the same advantages by means of a structure that would be called pBp because it would include a barrier layer between two electron-acceptor- type (p-type) layers.] The proposed modifications could make it practical to utilize nBn photodetectors in conjunction with readily available, compact thermoelectric coolers in diverse infrared- imaging applications that could include planetary exploration, industrial quality control, monitoring pollution, firefighting, law enforcement, and medical diagnosis.

P-type sp3-bonded BN/n-type Si heterodiode solar cell fabricated by laser-plasma synchronous CVD method

International Nuclear Information System (INIS)

Komatsu, Shojiro; Nagata, Takahiro; Chikyo, Toyohiro; Sato, Yuhei; Watanabe, Takayuki; Hirano, Daisuke; Takizawa, Takeo; Nakamura, Katsumitsu; Hashimoto, Takuya; Nakamura, Takuya; Koga, Kazunori; Shiratani, Masaharu; Yamamoto, Atsushi

2009-01-01

A heterojunction of p-type sp 3 -bonded boron nitride (BN) and n-type Si fabricated by laser-plasma synchronous chemical vapour deposition (CVD) showed excellent rectifying properties and proved to work as a solar cell with photovoltaic conversion efficiency of 1.76%. The BN film was deposited on an n-type Si (1 0 0) substrate by plasma CVD from B 2 H 6 + NH 3 + Ar while doping of Si into the BN film was induced by the simultaneous irradiation of an intense excimer laser with a pulse power of 490 mJ cm -2 , at a wavelength of 193 nm and at a repetition rate of 20 Hz. The source of dopant Si was supposed to be the Si substrate ablated at the initial stage of the film growth. The laser enhanced the doping (and/or diffusion) of Si into BN as well as the growth of sp 3 -bonded BN simultaneously in this method. P-type conduction of BN films was determined by the hot (thermoelectric) probe method. The BN/Si heterodiode with an essentially transparent p-type BN as a front layer is supposed to efficiently absorb light reaching the active region so as to potentially result in high efficiency.

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

International Nuclear Information System (INIS)

Petrenko, A.A.; Poplavsky, V.M.

1990-01-01

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

The theoretical possibility of reducing the doubling time in a fast-reactor by using heterogeneous configurations of various types of fuel

International Nuclear Information System (INIS)

Orlov, V.V.; Slesarev, I.S.; Zaritskij, S.M.; Subbotin, S.A.; Alekseev, P.N.; Zverkov, Yu.A.

1980-01-01

The authors have derived approximate expressions relating the doubling time of a fast reactor using various types of fuel simultaneously to the doubling time of traditional (homogeneous) reactors in which these types of fuel are used separately. These relationships afford a means of determining the conditions in which the use of various types of fuel can result in an improved doubling time. It was established that the use of heterogeneous compositions formed from assemblies of homogeneous systems gives a notable gain in doubling time over that of any of the original homogeneous systems if the doubling times were similar to each other. This gain is fairly large even in the case of BN reactors with high fuel volume fractions. The size of the gain depends on the degree of ''differentiation'' in the neutron and thermal properties of the components of the heterogeneous reactor. An optimum proportion has been found for the assemblies taken from the original homogeneous systems, governed primarily by the ratio of fuel densities. Estimates were made of the advantages of metallic oxide compositions over the traditional compositions used in large, fast reactors of the BN type. These estimates indicate that the former can be considered as alternative homogeneous compositions with carbide or nitride fuel as far as breeding characteristics are concerned. (author)

Engineering few-layer MoTe2 devices by Co/hBN tunnel contacts

Science.gov (United States)

Zhu, Mengjian; Luo, Wei; Wu, Nannan; Zhang, Xue-ao; Qin, Shiqiao

2018-04-01

2H phase Molybdenum ditelluride (MoTe2) is a layered two-dimensional (2D) semiconductor that has recently gained extensive attention for its intriguing properties, demonstrating great potential for nanoelectronics and optoelectronics. Optimizing the electric contacts to MoTe2 is a critical step for realizing high performance devices. Here, we demonstrate Co/hBN tunnel contacts to few-layer MoTe2. In sharp contrast to the p-type conduction of Co contacted MoTe2, Co/hBN tunnel contacted MoTe2 devices show clear n-type transport properties. Our first principles calculation reveals that the inserted few-layer hBN strongly interacts with Co and significantly reduces its work-function by ˜1.2 eV, while MoTe2 itself has a much weaker influence on the work-function of Co. This allows us to build MoTe2 diodes using the mixed Co/hBN and Co contact architecture, which can be switched from p-n type to n-p type by changing the gate-voltage, paving the way for engineering multi-functional devices based on atomically thin 2D semiconductors.

Advanced designs of VVER reactor plant

International Nuclear Information System (INIS)

Mokhov, V.A.

2010-01-01

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

Growth and characterization of thick cBN coatings on silicon and tool substrates

International Nuclear Information System (INIS)

Bewilogua, K.; Keunecke, M.; Weigel, K.; Wiemann, E.

2004-01-01

Recently some research groups have achieved progress in the deposition of cubic boron nitride (cBN) coatings with a thickness of 2 μm and more, which is necessary for cutting tool applications. In our laboratory, thick cBN coatings were sputter deposited on silicon substrates using a boron carbide target. Following a boron carbide interlayer (few 100 nm thick), a gradient layer with continuously increasing nitrogen content was prepared. After the cBN nucleation, the process parameters were modified for the cBN film growth to a thickness of more than 2 μm. However, the transfer of this technology to technically relevant substrates, like cemented carbide cutting inserts, required some further process modifications. At first, a titanium interlayer had to be deposited followed by a more than 1-μm-thick boron carbide layer. The next steps were identical to those on silicon substrates. The total coating thickness was in the range of 3 μm with a 0.5- to nearly 1-μm-thick cBN top layer. In spite of the enormous intrinsic stress, both the coatings on silicon and on cemented carbide exhibited a good adhesion and a prolonged stability in humid air. Oxidation experiments revealed a stability of the coating system on cemented carbide up to 700 deg. C and higher. Coated cutting inserts were tested in turning operations with different metallic workpiece materials. The test results will be compared to those of well-established cutting materials, like polycrystalline cubic boron nitride (PCBN) and oxide ceramics, considering the wear of coated tools

Activation and Radiation Damage Behaviour of Russian Structural Materials for Fusion Reactors in the Fission and Fusion Reactors

International Nuclear Information System (INIS)

Blokhin, A.; Demin, N.; Chernov, V.; Leonteva-Smirnova, M.; Potapenko, M.

2006-01-01

Various structural low (reduced) activated materials have been proposed as a candidate for the first walls-blankets of fusion reactors. One of the main problems connected with using these materials - to minimise the production of long-lived radionuclides from nuclear transmutations and to provide with good technological and functional properties. The selection of materials and their metallurgical and fabrication technologies for fusion reactor components is influenced by this factor. Accurate prediction of induced radioactivity is necessary for the development of the fusion reactor materials. Low activated V-Ti-Cr alloys and reduced activated ferritic-martensitic steels are a leading candidate material for fusion first wall and blanket applications. At the present time a range of compositions and an impurity level are still being investigated to better understand the sensitive of various functional and activation properties to small compositional variations and impurity level. For the two types of materials mentioned above (V-Ti-Cr alloys and 9-12 % Cr f/m steels) and manufactured in Russia (Russia technologies) the analysis of induced activity, hydrogen and helium-production as well as the accumulation of such elements as C, N, O, P, S, Zn and Sn as a function of irradiation time was performed. Materials '' were irradiated '' by fission (BN-600, BOR-60) and fusion (Russian DEMO-C Reactor Project) typical neutron spectra with neutron fluency up to 10 22 n/cm 2 and the cooling time up to 1000 years. The calculations of the transmutation of elements and the induced radioactivity were carried out using the FISPACT inventory code, and the different activation cross-section libraries like the ACDAM, FENDL-2/A and the decay data library FENDL-2/D. It was shown that the level of impurities controls a long-term behaviour of induced activity and contact dose rate for materials. From this analysis the concentration limits of impurities were obtained. The generation of gas

Application of safeguards design principles to the spent-fuel bundle counters for 600-MW CANDU reactors

International Nuclear Information System (INIS)

Stirling, A.J.; Allen, V.H.

1979-01-01

The irradiated fuel bundle counters for CANDU 600-MW reactors provide the IAEA with a secure and independent means of estimating the inventory of the spent-fuel storage bay at each inspection. Their function is straightforward - to count the bundles entering the storage area through the normal transfer ports. However, location, reliability, security and operating requirements make them highly ''intelligent'' instruments which have required a major development programme. Moreover, the bundle counters incorporate principles which apply to many unattended safeguards instruments. For example, concealing the operating status from potential diverters eases reliability specifications, continuous self-checking gives the inspector confidence in the readout, independence from continuous station services improves tamper-resistance, and the detailed data display provides tamper indication and a high level of credibility. Each irradiated fuel-bundle counter uses four Geiger counters to detect the passage of fuel bundles as they pass sequentially through the field-of-view. A microprocessor analyses the sequence of the Geiger counter signals and determines the number and direction of bundles transferred. The readout for IAEA inspectors includes both a tally and a printed log. The printer is also used to alert the inspector to abnormal fuel movements, tampering, Geiger counter failures and contamination of the fuel transfer mechanism. (author)

Corrosion Compatibility Studies on Inconel-600 in NP Decontamination Solution

Energy Technology Data Exchange (ETDEWEB)

Park, Sang Yoon; Jung, Jun Young; Won, Huijun; Choi, Wangkyu; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

It is well known that corrosion and contamination process in the primary cooling circuit of nuclear reactors are essentially interrelated: the contaminant isotopes are mostly corrosion products activated in the reactor core, and the contamination takes place on the out-core of Inconel-600 surface. This radionuclide uptake takes place up to the inner oxide layer and oxide/metal interface. So, it is necessary to remove inner oxide layer as well as outer oxide layer for excellent decontamination effects. The outer oxide layers are composed of Fe{sub 3}O{sub 4} and NiFe{sub 2}O{sub 4}. On the other hand, the inner oxide layers are composed of Cr{sub 2}O{sub 3}, (Ni{sub 1-x}Ni{sub x})(Cr{sub 1-y}Fe{sub y}){sub 2}O{sub 4}, and FeCr{sub 2}O{sub 4}. Because of chromium in the trivalent oxidation state which is difficult to dissolve, the oxide layer has an excellent protectiveness and become hard to be decontaminated. Alkaline permanganate (AP) or nitric permanganate (NP) oxidative phase has been used to dissolve the chromium-rich oxide. A disadvantage of AP process is the generation of a large volume of secondary waste. On the other hand, that of NP process is the high corrosion rate for Ni-base alloys. Therefore, for the safe use of oxidative phase in PWR system decontamination, it is necessary to reformulate the NP chemicals for decrease of corrosion rate. This study describes the corrosion compatibility on Inconel-600 and type 304 stainless steel in NP decontamination solution for PWR applications. To evaluate the general corrosion properties, weight change of NP treated specimens was measured. NP treated specimen surface was observed using optical microscope (OM) and scanning electron microscopy (SEM) for the evaluation of the localized corrosion. The effect of additives on the corrosion of the specimens was also evaluated. This study describes the corrosion compatibility on Inconel-600 and type 304 stainless steel in NP decontamination solution for PWR applications

Corrosion Compatibility Studies on Inconel-600 in NP Decontamination Solution

International Nuclear Information System (INIS)

Park, Sang Yoon; Jung, Jun Young; Won, Huijun; Choi, Wangkyu; Moon, Jeikwon

2013-01-01

It is well known that corrosion and contamination process in the primary cooling circuit of nuclear reactors are essentially interrelated: the contaminant isotopes are mostly corrosion products activated in the reactor core, and the contamination takes place on the out-core of Inconel-600 surface. This radionuclide uptake takes place up to the inner oxide layer and oxide/metal interface. So, it is necessary to remove inner oxide layer as well as outer oxide layer for excellent decontamination effects. The outer oxide layers are composed of Fe 3 O 4 and NiFe 2 O 4 . On the other hand, the inner oxide layers are composed of Cr 2 O 3 , (Ni 1-x Ni x )(Cr 1-y Fe y ) 2 O 4 , and FeCr 2 O 4 . Because of chromium in the trivalent oxidation state which is difficult to dissolve, the oxide layer has an excellent protectiveness and become hard to be decontaminated. Alkaline permanganate (AP) or nitric permanganate (NP) oxidative phase has been used to dissolve the chromium-rich oxide. A disadvantage of AP process is the generation of a large volume of secondary waste. On the other hand, that of NP process is the high corrosion rate for Ni-base alloys. Therefore, for the safe use of oxidative phase in PWR system decontamination, it is necessary to reformulate the NP chemicals for decrease of corrosion rate. This study describes the corrosion compatibility on Inconel-600 and type 304 stainless steel in NP decontamination solution for PWR applications. To evaluate the general corrosion properties, weight change of NP treated specimens was measured. NP treated specimen surface was observed using optical microscope (OM) and scanning electron microscopy (SEM) for the evaluation of the localized corrosion. The effect of additives on the corrosion of the specimens was also evaluated. This study describes the corrosion compatibility on Inconel-600 and type 304 stainless steel in NP decontamination solution for PWR applications. It is revealed that Inconel-600 specimen is more

Preparation of transparent BN films with superhydrophobic surface

International Nuclear Information System (INIS)

Li Guoxing; Liu Yi; Wang Bo; Song Xuemei; Li Er; Yan Hui

2008-01-01

A novel approach was investigated to obtain the superhydrophobicity on surfaces of boron nitride films. In this method boron nitride films were deposited firstly on Si(1 0 0) and quartz substrate using a radio frequency (RF) magnetron sputtering system, and then using CF 4 plasma treatment, the topmost surface area can be modified systematically. The results have shown that the water contact angle on such surfaces can be tuned from 67 deg. to 159 deg. The films were observed to be uniform. The surfaces of films consist of micro-features, which were confirmed by Atomic Force Micrograph. The chemical bond states of the films were determined by Fourier Transform Infrared (FTIR) Spectroscopy, which indicate the dominance of B-N binding. According to the X-ray Photoelectron Spectroscopy analysis, the surface of film is mainly in BN phase. The micro-feature induced surface roughness is responsible for the observed superhydrophobic nature. The water contact angles measured on these surfaces can be modeled by the Cassie's formulation

Effects of epitaxial structure and processing on electrical characteristics of InAs-based nBn infrared detectors

Science.gov (United States)

Du, X.; Savich, G. R.; Marozas, B. T.; Wicks, G. W.

2017-02-01

The conventional processing of the III-V nBn photodetectors defines mesa devices by etching the contact n-layer and stopping immediately above the barrier, i.e., a shallow etch. This processing enables great suppression of surface leakage currents without having to explore surface passivation techniques. However, devices that are made with this processing scheme are subject to lateral diffusion currents. To address the lateral diffusion current, we compare the effects of different processing approaches and epitaxial structures of nBn detectors. The conventional solution for eliminating lateral diffusion current, a deep etch through the barrier and the absorber, creates increased dark currents and an increased device failure rate. To avoid deep etch processing, a new device structure is proposed, the inverted-nBn structure. By comparing with the conventional nBn structure, the results show that the lateral diffusion current is effectively eliminated in the inverted-nBn structure without elevating the dark currents.

Wear resistance and microstructural properties of Ni–Al/h-BN/WC–Co coatings deposited using plasma spraying

International Nuclear Information System (INIS)

Hsiao, W.T.; Su, C.Y.; Huang, T.S.; Liao, W.H.

2013-01-01

Hexagonal boron nitride (h-BN) and tungsten carbide cobalt (WC–Co) were added to nickel aluminum alloy (Ni–Al) and deposited as plasma sprayed coatings to improve their tribological properties. The microstructure of the coatings was analyzed using a scanning electron microscope (SEM). Following wear test, the worn surface morphologies of the coatings were analyzed using a SEM to identify their fracture modes. The results of this study demonstrate that the addition of h-BN and WC–Co improved the properties of the coatings. Ni–Al/h-BN/WC–Co coatings with high hardness and favorable lubrication properties were deposited. - Highlights: • We mixed Ni–Al, h-BN and WC–Co powders and deposited them as composite coatings. • Adding WC–Co was found to increase the hardness and reduce the wear volume loss. • Adding h-BN was found to decrease the hardness and reduce the friction coefficient. • This composite coating was shown to have improved wear properties at 850 °C

Effect of an in-plane ligand on the electronic structures of bromo-bridged nano-wire Ni-Pd mixed-metal complexes, [Ni(1-x)Pd(x)(bn)2Br]Br2 (bn = 2S,3S-diaminobutane).

Science.gov (United States)

Sasaki, Mari; Wu, Hashen; Kawakami, Daisuke; Takaishi, Shinya; Kajiwara, Takashi; Miyasaka, Hitoshi; Breedlove, Brian K; Yamashita, Masahiro; Kishida, Hideo; Matsuzaki, Hiroyuki; Okamoto, Hiroshi; Tanaka, Hisaaki; Kuroda, Shinichi

2009-08-03

Single crystals of quasi-one-dimensional bromo-bridged Ni-Pd mixed-metal complexes with 2S,3S-diaminobutane (bn) as an in-plane ligand, [Ni(1-x)Pd(x)(bn)(2)Br]Br(2), were obtained by using an electrochemical oxidation method involving mixed methanol/2-propanol (1:1) solutions containing different ratios of [Ni(II)(bn)(2)]Br(2) and [Pd(II)(bn)(2)]Br(2). To investigate the competition between the electron-correlation of the Ni(III) states, or Mott-Hubbard states (MH), and the electron-phonon interaction of the Pd(II)-Pd(IV) mixed valence states, or charge-density-wave states (CDW), in the Ni-Pd mixed-metal compounds, X-ray structure analyses, X-ray oscillation photograph, and Raman, IR, ESR, and single-crystal reflectance spectra were analyzed. In addition, the local electronic structures of Ni-Pd mixed-metal single crystals were directly investigated by using scanning tunneling microscopy (STM) at room temperature and ambient pressure. The oxidation states of [Ni(1-x)Pd(x)(bn)(2)Br]Br(2) changed from a M(II)-M(IV) mixed valence state to a M(III) MH state at a critical mixing ratio (x(c)) of approximately 0.8, which is lower than that of [Ni(1-x)Pd(x)(chxn)(2)Br]Br(2) (chxn = 1R,2R-diaminocyclohexane) (x(c) approximately 0.9) reported previously. The lower value of x(c) for [Ni(1-x)Pd(x)(bn)(2)Br]Br(2) can be explained by the difference in their CDW dimensionalities because the three-dimensional CDW ordering in [Pd(bn)(2)Br]Br(2) observed by using X-ray diffuse scattering stabilizes the Pd(II)-Pd(IV) mixed valence state more than two-dimensional CDW ordering in [Pd(chxn)(2)Br]Br(2) does, which has been reported previously.

Search for muonium states in BN, WS[sub 2] and carbon nanotubes

Energy Technology Data Exchange (ETDEWEB)

Ansaldo, E J [TRIUMF, Vancouver (Canada) Univ. of Saskatchewan, Saskatoon (Canada)

1994-07-01

A sizable missing fraction was found for semiconductors (cubic) BN and (hexagonal) WS[sub 2]. A repolarization measurement at room temperature yielded a high hyperfine frequency for the dominant muonium signal in BN. The missing fraction in carbon nanotubes with average 20 nm outer diameter was less than 4% at temperatures above 5 K, with very small relaxation in transverse and longitudinal fields, indicating that such tubulenes are microscopically (semi-)metals or small-gap semiconductors, non-magnetic and non-superconducting. (orig.)

Controlling the orientation of nucleobases by dipole moment interaction with graphene/h-BN interfaces

KAUST Repository

Vovusha, Hakkim; Amorim, Rodrigo G.; Scheicher, Ralph H.; Sanyal, Biplab

2018-01-01

The interfaces in 2D hybrids of graphene and h-BN provide interesting possibilities of adsorbing and manipulating atomic and molecular entities. In this paper, with the aid of density functional theory, we demonstrate the adsorption characteristics of DNA nucleobases at different interfaces of 2D hybrid nanoflakes of graphene and h-BN. The interfaces provide stronger binding to the nucleobases in comparison to pure graphene and h-BN nanoflakes. It is also revealed that the individual dipole moments of the nucleobases and nanoflakes dictate the orientation of the nucleobases at the interfaces of the hybrid structures. The results of our study point towards a possible route to selectively control the orientation of individual molecules in biosensors.

Controlling the orientation of nucleobases by dipole moment interaction with graphene/h-BN interfaces

KAUST Repository

Vovusha, Hakkim

2018-02-08

The interfaces in 2D hybrids of graphene and h-BN provide interesting possibilities of adsorbing and manipulating atomic and molecular entities. In this paper, with the aid of density functional theory, we demonstrate the adsorption characteristics of DNA nucleobases at different interfaces of 2D hybrid nanoflakes of graphene and h-BN. The interfaces provide stronger binding to the nucleobases in comparison to pure graphene and h-BN nanoflakes. It is also revealed that the individual dipole moments of the nucleobases and nanoflakes dictate the orientation of the nucleobases at the interfaces of the hybrid structures. The results of our study point towards a possible route to selectively control the orientation of individual molecules in biosensors.

Structure of boron clusters revisited, Bn with n = 14-20

Science.gov (United States)

Tai, Truong Ba; Tam, Nguyen Minh; Nguyen, Minh Tho

2012-03-01

We reinvestigate the structures of neutral boron clusters Bn, with n = 14-20. G3B3 calculations confirm that a transition between 2D and 3D shape occurs at B20, which has a tubular form. In disagreement with Boustani et al. (Phys. Rev. B, 83 (2011) 193405), we find a planar B19 cluster. Standard heats of formation are obtained and used to evaluate the clusters stability. The average binding energy tends to increase with increasing size toward a limit. Higher stability is found B14, B16, B18 and B20. All Bn have negative NICS-values. The bonding nature and electron delocalization of B20 are re-examined using CMO and LOL.

Electron scattering in graphene by defects in underlying h-BN layer: First-principles transport calculations

Science.gov (United States)

Kaneko, Tomoaki; Ohno, Takahisa

2018-03-01

We investigate the electronic structure and the transport properties of graphene adsorbed onto h-BN with carbon impurities or atomic vacancies using density functional theory and the non-equilibrium Green's function method. We find that the transport properties are degraded due to carrier doping and scattering off of localized defect states in h-BN. When graphene is doped by introducing defects in h-BN, the transmission spectra become asymmetric owing to the reduction of the electronic density of states, which contributes significantly to the degradation of graphene transport properties as compared with the effect of defect levels.

Investigation of band structure and electrochemical properties of h-BN/rGO composites for asymmetric supercapacitor applications

Energy Technology Data Exchange (ETDEWEB)

Saha, Sanjit; Jana, Milan; Samanta, Pranab; Murmu, Naresh C. [Surface Engineering & Tribology Division, CSIR-Central Mechanical Engineering Research Institute, Durgapur, 713209 (India); Academy of Scientific and Innovative Research (AcSIR), CSIR-CMERI Campus, Durgapur, 713209 (India); Kim, Nam H. [Advanced Materials Institute of BIN Convergence Technology (BK21 Plus Global), Dept. of BIN Convergence Technology, Chonbuk National University, Jeonju, Jeonbuk, 54896 (Korea, Republic of); Kuila, Tapas, E-mail: tkuila@gmail.com [Surface Engineering & Tribology Division, CSIR-Central Mechanical Engineering Research Institute, Durgapur, 713209 (India); Academy of Scientific and Innovative Research (AcSIR), CSIR-CMERI Campus, Durgapur, 713209 (India); Lee, Joong H., E-mail: jhl@jbnu.ac.kr [Advanced Materials Institute of BIN Convergence Technology (BK21 Plus Global), Dept. of BIN Convergence Technology, Chonbuk National University, Jeonju, Jeonbuk, 54896 (Korea, Republic of); Carbon Composite Research Centre, Department of Polymer & Nanoscience and Technology, Chonbuk National University, Jeonju, Jeonbuk, 54896 (Korea, Republic of)

2017-04-01

The effect of different content of graphene oxide (GO) on the electrical and electrochemical property of h-BN/reduced GO (rGO) hetero-structure is investigated elaborately. The increasing amount of rGO within the h-BN moiety plays fascinating role by reducing the electronic work function while increasing the density of state of the electrode. Furthermore, different h-BN/rGO architecture shows different potential window and the transition from pseudocapacitance to electrochemical double layer capacitance (EDLC) is observed with increasing π-conjugation of C atoms. The rod like h-BN is aligned as sheet while forming super-lattice with rGO. Transmission electron microscopy images show crystalline morphology of the hetero-structure super-lattice. The valance band and Mott-Shotky relationship determined from Mott-Shotky X-ray photoelectron spectroscopy shows that the electronic band structure of super-lattice is improved as compared to the insulating h-BN. The h-BN/rGO super-lattice provides high specific capacitance of ∼960 F g{sup −1}. An asymmetric device configured with h-BN/rGO super-lattice and B, N doped rGO shows very high energy and power density of 73 W h kg{sup −1} and 14,000 W kg{sup −1}, respectively. Furthermore, very low relaxation time constant of ∼1.6 ms and high stability (∼80%) after 10,000 charge-discharge cycles ensure the h-BN/rGO super-lattice as potential materials for the next generation energy storage applications. - Highlights: • Band gap energy of boron nitride decreased with increasing graphene oxide content. • Graphene oxide effectively affected the charge storage mechanism of the composite. • Morphology of boron nitride changed from rod to sheet while forming superlattice. • Highly conducting superlattice showed excellent supercapacitor performance. • Asymmetric device exhibited long stability with high energy and power density.

Germany unveils €18bn research plan

Science.gov (United States)

Banks, Michael

2009-07-01

The German government has unveiled an ambitious plan to inject a total of €18bn into teaching and research over the next decade. The German chancellor Angela Merkel, who has a degree in physics, announced that she was releasing the funds despite concerns from her social-democrat coalition partners that financing the package could be difficult in the economic downturn.

Reactor types for the future

International Nuclear Information System (INIS)

Hall, A.C.

1990-01-01

The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

Reactor types for the future

Energy Technology Data Exchange (ETDEWEB)

Hall, A C [PWR Power Projects Ltd., Knutsford, Cheshire (United Kingdom)

1990-06-01

The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

Abstracts of 2. international conference C-BN and diamond crystallization under reduced pressure

International Nuclear Information System (INIS)

1995-01-01

The important problem and the last advanced one from the view point of electronic materials sciences is the new A III B V compounds creation and investigation of their properties. This domain was the main subject of the 2. International Conference on C-BN and diamond crystallization under reduced pressure. The conference has been divided into 8 sessions. They were: opening address, c-BN, new materials, posters, diamond, applications, posters

PREFACE: Ultrathin layers of graphene, h-BN and other honeycomb structures Ultrathin layers of graphene, h-BN and other honeycomb structures

Science.gov (United States)

Geber, Thomas; Oshima, Chuhei

2012-08-01

Since ancient times, pure carbon materials have been familiar in human society—not only diamonds in jewellery and graphite in pencils, but also charcoal and coal which have been used for centuries as fuel for living and industry. Carbon fibers are stronger, tougher and lighter than steel and increase material efficiency because of their lower weight. Today, carbon fibers and related composite materials are used to make the frames of bicycles, cars and even airplane parts. The two-dimensional allotrope, now called graphene, is just a single layer of carbon atoms, locked together in a strongly bonded honeycomb lattice. In plane, graphene is stiffer than diamond, but out-of-plane it is soft, like rubber. It is virtually invisible, may conduct electricity (heat) better than copper and weighs next to nothing. Carbon compounds with two carbon atoms as a base, such as graphene, graphite or diamond, have isoelectronic sister compounds made of boron-nitrogen pairs: hexagonal and cubic boron nitride, with almost the same lattice constant. Although the two 2D sisters, graphene and h-BN, have the same number of valence electrons, their electronic properties are very different: freestanding h-BN is an insulator, while charge carriers in graphene are highly mobile. The past ten years have seen a great expansion in studies of single-layer and few-layer graphene. This activity has been concerned with the π electron transport in graphene, in electric and magnetic fields. More than 30 years ago, however, single-layer graphene and h-BN on solid surfaces were widely investigated. It was noted that they drastically changed the chemical reactivity of surfaces, and they were known to 'poison' heterogeneous catalysts, to passivate surfaces, to prevent oxidation of surfaces and to act as surfactants. Also, it was realized that the controlled growth of h-BN and graphene on substrates yields the formation of mismatch driven superstructures with peculiar template functionality on the

Core concept of fast power reactor with zero sodium void reactivity

International Nuclear Information System (INIS)

Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

1991-01-01

The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

Raman enhancement effect on two-dimensional layered materials: graphene, h-BN and MoS2.

Science.gov (United States)

Ling, Xi; Fang, Wenjing; Lee, Yi-Hsien; Araujo, Paulo T; Zhang, Xu; Rodriguez-Nieva, Joaquin F; Lin, Yuxuan; Zhang, Jin; Kong, Jing; Dresselhaus, Mildred S

2014-06-11

Realizing Raman enhancement on a flat surface has become increasingly attractive after the discovery of graphene-enhanced Raman scattering (GERS). Two-dimensional (2D) layered materials, exhibiting a flat surface without dangling bonds, were thought to be strong candidates for both fundamental studies of this Raman enhancement effect and its extension to meet practical applications requirements. Here, we study the Raman enhancement effect on graphene, hexagonal boron nitride (h-BN), and molybdenum disulfide (MoS2), by using the copper phthalocyanine (CuPc) molecule as a probe. This molecule can sit on these layered materials in a face-on configuration. However, it is found that the Raman enhancement effect, which is observable on graphene, hBN, and MoS2, has different enhancement factors for the different vibrational modes of CuPc, depending strongly on the surfaces. Higher-frequency phonon modes of CuPc (such as those at 1342, 1452, 1531 cm(-1)) are enhanced more strongly on graphene than that on h-BN, while the lower frequency phonon modes of CuPc (such as those at 682, 749, 1142, 1185 cm(-1)) are enhanced more strongly on h-BN than that on graphene. MoS2 demonstrated the weakest Raman enhancement effect as a substrate among these three 2D materials. These differences are attributed to the different enhancement mechanisms related to the different electronic properties and chemical bonds exhibited by the three substrates: (1) graphene is zero-gap semiconductor and has a nonpolar C-C bond, which induces charge transfer (2) h-BN is insulating and has a strong B-N bond, while (3) MoS2 is semiconducting with the sulfur atoms on the surface and has a polar covalent bond (Mo-S) with the polarity in the vertical direction to the surface. Therefore, the different Raman enhancement mechanisms differ for each material: (1) charge transfer may occur for graphene; (2) strong dipole-dipole coupling may occur for h-BN, and (3) both charge transfer and dipole-dipole coupling may

Studying the processes of sodium-water interaction in the BOR-60 reactor micromodule steam generator

International Nuclear Information System (INIS)

Tsykanov, V.A.; Antipin, G.K.; Borisov, V.V.

1981-01-01

Main results of experimental studies of emergency regimes of micromodule steam generator (MSG) at small and big leaks of water into sodium, realized using the 30 MW MSG, operating in the BOR-o0 reactor, are considered. The aims of the study are as follows: the modelling of macroleak in ''Nadja'' steam generator for the BN-350 reactor; testing the conceptions of alarm signalling and MSG protection; testing under real conditions of new perspective systems of leak detection; gaining the experimence and development of the ways to eliminate the consequences of accident caused by big water leak into sodium; accumulation of knowledge on restoration of MSG operating ability after accident; experimental test of calculational techniques for big leak accidents to use them in future for calculational studies of similar situations at other reactors equipped with sodium-water steam generators; refinement of characteristics of hydrodynamic and thermal effects interaction zone for big leak in real circuit during the plant operation. A series of experiments with the imitation of water leak into sodium by means of argon and steam supply through injection devices, located before the steam superheater module of one of the sections and between evaporator module of the same section, is conducted. The range of steam flow rate is 0.02-0.45 g/s. Duration of steam supply is 100-400 s. A conclusion is made that the results obtained can be used for steam generator of the BN-350 reactor [ru

Multiple-walled BN nanotubes obtained with a mechanical alloying technique

International Nuclear Information System (INIS)

Rosas, G.; Sistos, J.; Ascencio, J.A.; Medina, A.; Perez, R.

2005-01-01

An experimental method to obtain multiple-walled nanotubes of BN using low energy is presented. The method is based on the use of mechanical alloying techniques with elemental boron powders and nitrogen gas mixed in an autoclave at room temperature. The chemical and structural characteristics of the multiple-walled nanotubes were obtained using different techniques, such as X-ray diffraction, transmission electron microscopy, EELS microanalysis, high-resolution electron microscopy images and theoretical simulations based on the multisliced approach of the electron diffraction theory. This investigation clearly illustrates the production of multiple-wall BN nanotubes at room temperature. These results open up a new kind of synthesis method with low expense and important perspectives for use in large-quantity production. (orig.)

The next generation of power reactors - safety characteristics

International Nuclear Information System (INIS)

Modro, S.M.

1995-01-01

The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

Electronic structure and STM images simulation of defects on hBN/ black-phosphorene heterostructures: A theoretical study

Science.gov (United States)

Ospina, D. A.; Cisternas, E.; Duque, C. A.; Correa, J. D.

2018-03-01

By first principles calculations which include van der Waals interactions, we studied the electronic structure of hexagonal boron-nitride/black-phosphorene heterostructures (hBN/BP). In particular the role of several kind of defects on the electronic properties of black-phosphorene monolayer and hBN/BP heterostructure was analyzed. The defects under consideration were single and double vacancies, as well Stone-Wale type defects, all of them present in the phosphorene layer. In this way, we found that the electronic structure of the hBN/BP is modified according the type of defect that is introduced. As a remarkable feature, our results show occupied states at the Fermi Level introduced by a single vacancy in the energy gap of the hBN/BP heterostructure. Additionally, we performed simulations of scanning tunneling microscopy images. These simulations show that is possible to discriminate the kind of defect even when the black-phosphorene monolayer is part of the heterostructure hBN/BP. Our results may help to discriminate among several kind of defects during experimental characterization of these novel materials.

The electrical and thermal transport properties of hybrid zigzag graphene-BN nanoribbons

Science.gov (United States)

Gao, Song; Lu, Wei; Zheng, Guo-Hui; Jia, Yalei; Ke, San-Huang

2017-06-01

The electron and phonon transport in hybrid graphene-BN zigzag nanoribbons are investigated by the nonequilibrium Green’s function method combined with density functional theory calculations. A 100% spin-polarized electron transport in a large energy window around the Fermi level is found and this behavior is independent of the ribbon width as long as there contain 3 zigzag carbon chains. The phonon transport calculations show that the ratio of C-chain number to BN-chain number will modify the thermal conductance of the hybrid nanoribbon in a complicated manner.

The electrical and thermal transport properties of hybrid zigzag graphene-BN nanoribbons

International Nuclear Information System (INIS)

Gao, Song; Lu, Wei; Zheng, Guo-Hui; Jia, Yalei; Ke, San-Huang

2017-01-01

The electron and phonon transport in hybrid graphene-BN zigzag nanoribbons are investigated by the nonequilibrium Green’s function method combined with density functional theory calculations. A 100% spin-polarized electron transport in a large energy window around the Fermi level is found and this behavior is independent of the ribbon width as long as there contain 3 zigzag carbon chains. The phonon transport calculations show that the ratio of C-chain number to BN-chain number will modify the thermal conductance of the hybrid nanoribbon in a complicated manner. (paper)

Study of vibrational modes and specific heat of wurtzite phase of BN

Energy Technology Data Exchange (ETDEWEB)

Singh, Daljit, E-mail: daljit.jt@gmail.com; Sinha, M. M. [Department of Physics, SLIET, Longowal (India)

2016-05-06

In these days of nanotechnology the materials like BN is of utmost importance as in hexagonal phase it is among hardest materials. The phonon mode study of the materials is most important factor to find structural and thermodynamcal properties. To study the phonons de launey angular force (DAF) constant model is best suited as it involves many particle interactions. Therefore in this presentation we have studied the lattice dynamical properties and specific heat of BN in wurtzite phase using DAF model. The obtained results are in excellent agreement with existing results.

Study of vibrational modes and specific heat of wurtzite phase of BN

International Nuclear Information System (INIS)

Singh, Daljit; Sinha, M. M.

2016-01-01

In these days of nanotechnology the materials like BN is of utmost importance as in hexagonal phase it is among hardest materials. The phonon mode study of the materials is most important factor to find structural and thermodynamcal properties. To study the phonons de launey angular force (DAF) constant model is best suited as it involves many particle interactions. Therefore in this presentation we have studied the lattice dynamical properties and specific heat of BN in wurtzite phase using DAF model. The obtained results are in excellent agreement with existing results.

Modeling of grain boundary stresses in Alloy 600

Energy Technology Data Exchange (ETDEWEB)

Kozaczek, K.J. [Oak Ridge National Lab., TN (United States); Sinharoy, A.; Ruud, C.O. [Pennsylvania State Univ., University Park, PA (United States); Mcllree, A.R. [Electric Power Research Inst., Palo Alto, CA (United States)

1995-04-01

Corrosive environments combined with high stress levels and susceptible microstructures can cause intergranular stress corrosion cracking (IGSCC) of Alloy 600 components on both primary and secondary sides of pressurized water reactors. One factor affecting the IGSCC is intergranular carbide precipitation controlled by heat treatment of Alloy 600. This study is concerned with analysis of elastic stress fields in vicinity of M{sub 7}C{sub 3} and M{sub 23}C{sub 6} carbides precipitated in the matrix and at a grain boundary triple point. The local stress concentration which can lead to IGSCC initiation was studied using a two-dimensional finite element model. The intergranular precipitates are more effective stress raisers than the intragranular precipitates. The combination of the elastic property mismatch and the precipitate shape can result in a local stress field substantially different than the macroscopic stress. The maximum local stresses in the vicinity of the intergranular precipitate were almost twice as high as the applied stress.

Turning into carbonate the residual sodium left in BN-350 circuits may alleviate concerns over their long term safe confinement

International Nuclear Information System (INIS)

Rahmani, L

2000-01-01

After the coolant is drained from the reactor vessel and from the primary and secondary circuits of the BN-350 nuclear power plant, what sodium is left in ponds and films may amount to hundreds of kilograms. For the long term safe storage period which is to follow, preliminary safety analyses (e.g. derived from those made for French sodium cooled reactors) might show that the risks incurred through loss of leaktightness are significant. The ingress of moisture into the circuits would generate, by reaction with the sodium, two undesirable products : sodium hydroxide and hydrogene. Even when considering that water would enter the circuits progressively, so that the heat of the reaction does not give rise to over-pressure, some main risk factors remain. The most promising solution to this challenge appears to be the carbonation of the sodium residues, by progressive diffusion of an appropriate association of carbon dioxyde and water vapour through the inert gaseous medium which fills the circuits. The desired product is porous sodium hydrogenocarbonate

BnDGAT1s Function Similarly in Oil Deposition and Are Expressed with Uniform Patterns in Tissues of Brassica napus

Directory of Open Access Journals (Sweden)

Cuizhu Zhao

2017-12-01

Full Text Available As an allotetraploid oilcrop, Brassica napus contains four duplicated Acyl-CoA:diacylglycerol acyltransferase 1 (DGAT1 genes, which catalyze one of the rate-limiting steps in triacylglycerol (TAG biosynthesis in plants. While all four BnDGAT1s have been expressed functionally in yeast, their expression patterns in different germplasms and tissues and also consequent contribution to seed oil accumulation in planta remain to be elucidated. In this study, the coding regions of the four BnDGAT1s were expressed in an Arabidopsis dgat1 mutant. All four BnDGAT1s showed similar effects on oil content and fatty acid composition, a result which is different from that observed in previous studies of their expression in yeast. Expression patterns of BnDGAT1s were analyzed in developing seeds of 34 B. napus inbred lines and in different tissues of 14 lines. Different expression patterns were observed for the four BnDGAT1s, which suggests that they express independently or randomly in different germplasm sources. Higher expression of BnDGAT1s was correlated with higher seed oil content lines. Tissue-specific analyses showed that the BnDGAT1s were expressed in a uniform pattern in different tissues. Our results suggest that it is important to maintain expression of the four BnDGAT1s for maximum return on oil content.

Technical meeting on decommissioning of fast reactors after sodium draining. Working material

International Nuclear Information System (INIS)

2005-01-01

The objective of the technical meeting was to provide a forum for in-depth scientific and technical exchange on topics related to the decommissioning experience with fast reactors, in particular with regard to the decommissioning of components after sodium draining. Accordingly, the scope of the meeting covers the review and analyses of the experience gained from the decommissioning of both active sodium loops and sodium cooled fast reactors (e.g., KNK II, Superphenix, RAPSODIE, EBR-II, FERMI, BN-350, BR-10). It is expected that the outcome of the meeting will contribute to the Agency initiative to preserve fast reactor data and knowledge. The main focus of the technical meeting was given on the decommissioning of both active loop and reactor components (e.g., the primary vessel of a sodium-cooled reactor) that have been drained of sodium, but that still conserve some residual amounts of sodium (e.g., films covering the entire surface of the component, or particular sodium heels that cannot be drained)

BnEPFL6, an EPIDERMAL PATTERNING FACTOR-LIKE (EPFL) secreted peptide gene, is required for filament elongation in Brassica napus.

Science.gov (United States)

Huang, Yi; Tao, Zhangsheng; Liu, Qiong; Wang, Xinfa; Yu, Jingyin; Liu, Guihua; Wang, Hanzhong

2014-07-01

Inflorescence architecture, pedicel length and stomata patterning in Arabidopsis thaliana are specified by inter-tissue communication mediated by ERECTA and its signaling ligands in the EPIDERMAL PATTERNING FACTOR-LIKE (EPFL) family of secreted cysteine-rich peptides. Here, we identified and characterized BnEPFL6 from Brassica napus. Heterologous expression of this gene under the double enhanced CaMV promoter (D35S) in Arabidopsis resulted in shortened stamen filaments, filaments degradation, and reduced filament cell size that displayed down-regulated expression of AHK2, in which phenotypic variation of ahk2-1 mutant presented highly consistent with that of BnEPFL6 transgenic lines. Especially, the expression level of BnEPFL6 in the shortened filaments of four B. napus male sterile lines (98A, 86A, SA, and Z11A) was similar to that of BnEPFL6 in the transgenic Arabidopsis lines. The activity of pBnEPFL6.2::GUS was intensive in the filaments of transgenic lines. These observations reveal that BnEPFL6 plays an important role in filament elongation and may also affect organ morphology and floral organ specification via a BnEPFL6-mediated cascade.

Boosting the adsorption performance of BN nanosheet as an anode of Na-ion batteries: DFT studies

Energy Technology Data Exchange (ETDEWEB)

Hosseinian, A. [Department of Engineering Science, College of Engineering, University of Tehran, P.O. Box 11365-4563, Tehran (Iran, Islamic Republic of); Soleimani-amiri, S. [Department of Chemistry, Karaj Branch, Islamic Azad University, Karaj (Iran, Islamic Republic of); Arshadi, S., E-mail: chemistry_arshadi@pnu.ac.ir [Department of Chemistry, Payame Noor University, Tehran (Iran, Islamic Republic of); Vessally, E. [Department of Chemistry, Payame Noor University, Tehran (Iran, Islamic Republic of); Edjlali, L. [Department of Chemistry, Tabriz Branch, Islamic Azad University, Tabriz (Iran, Islamic Republic of)

2017-06-28

Despite the high advance in the Li-ion battery technology, there exist great concerns about its lifetime, safety, cost, and low-temperature performance. It is expected that the Li-ion batteries may be replaced by Na-ion batteries (NIB) because of the low cost, nontoxicity, and wide availability of sodium. Here, we investigated the potential application of BN nanosheets in anode of NIBs by means of density functional theory calculation and introduced a strategy to increase their performance. It was shown that the Na and Na{sup +} are mainly adsorbed on the center of a hexagonal ring of BN sheet with adsorption energies of −0.08 and −33.7 kcal/mol, respectively. Replacing three N atoms of the hexagonal ring with larger P atoms significantly increases the performance of the sheet as an anode of a NIB but the replacement of B by Al decreases the performance. The initial cell voltage of LIB is increased by about 0.67 V after the P-doping which causes a high storage performance with long discharge time. The results are discussed based on the energetic, structural, orbital, charge transfer and electronic properties and provide guidelines to build better high-capacity anode materials for NIBs. - Highlights: • Potential use of BN sheet as anode in Na-ion batteries (NIB) is studied by DFT. • The replacement of B by Al decreases the performance. • The cell voltage of LIB is increased by about 0.67 V after by P-doping. • The order of performance is P-BN > BN >> Al-BN.

Creep of Sylramic-iBN Fiber Tows at Elevated Temperature in Air and in Silicic Acid-Saturated Steam

Science.gov (United States)

2015-06-01

CREEP OF SYLRAMIC-iBN FIBER TOWS AT ELEVATED TEMPERATURE IN AIR AND IN SILICIC ACID-SATURATED STEAM ...protection in the United States. AFIT-ENY-15-J-46 CREEP OF SYLRAMIC-iBN FIBER TOWS AT ELEVATED TEMPERATURE IN AIR AND IN SILICIC ACID-SATURATED STEAM ...DISTRIBUTION UNLIMITED. AFIT-ENY-15-J-46 CREEP OF SYLRAMIC-iBN FIBER TOWS AT ELEVATED TEMPERATURE IN AIR AND IN SILICIC ACID-SATURATED STEAM

Cleanup Verification Package for the 600-259 Waste Site

Energy Technology Data Exchange (ETDEWEB)

J. M. Capron

2006-02-09

This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste.

Cleanup Verification Package for the 600-259 Waste Site

International Nuclear Information System (INIS)

Capron, J.M.

2006-01-01

This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste

Water-chemical regime of a fast reactor ower complex

International Nuclear Information System (INIS)

Musikhin, R.N.; Piskunov, E.M.; Samarkin, A.A.; Yurchenko, D.S.

1983-01-01

Some peculiarities of water-chemical regime of a power compleX in Shevchenko are considered. The complex comprises a desalination unit, a gas-masout heating-and-power plant and the BN-350 reactor. The compleX is used for the production of electric and thermal energy and fresh water. The power complex peculiarity is the utilization of disalinated seawater in a technological cycle along with highly mineralized seawater with a total salt content of 13.5 g/l (for cooling) in heat exchanges. A regime of ammoniacal correction of feed water was used as a basic water-chemical regime in the initial period of the BN-350 steam generator operation. Deposits composed mainly of iron oxide slime were observed on steam generator surfaces during the operation under these conditions. A conclusion is made that the regime with chelating agent providing steam generator safe operation without chemical cleaning is the most expedient one

Bandgap renormalization and work function tuning in MoSe2/hBN/Ru(0001) heterostructures.

Science.gov (United States)

Zhang, Qiang; Chen, Yuxuan; Zhang, Chendong; Pan, Chi-Ruei; Chou, Mei-Yin; Zeng, Changgan; Shih, Chih-Kang

2016-12-14

The van der Waals interaction in vertical heterostructures made of two-dimensional (2D) materials relaxes the requirement of lattice matching, therefore enabling great design flexibility to tailor novel 2D electronic systems. Here we report the successful growth of MoSe 2 on single-layer hexagonal boron nitride (hBN) on the Ru(0001) substrate using molecular beam epitaxy. Using scanning tunnelling microscopy and spectroscopy, we found that the quasi-particle bandgap of MoSe 2 on hBN/Ru is about 0.25 eV smaller than those on graphene or graphite substrates. We attribute this result to the strong interaction between hBN/Ru, which causes residual metallic screening from the substrate. In addition, the electronic structure and the work function of MoSe 2 are modulated electrostatically with an amplitude of ∼0.13 eV. Most interestingly, this electrostatic modulation is spatially in phase with the Moiré pattern of hBN on Ru(0001) whose surface also exhibits a work function modulation of the same amplitude.

Decommissioning of the CANDU-PHW reactor

International Nuclear Information System (INIS)

Unsworth, G.N.

1977-04-01

This report contains the results of a study of various aspects of decommissioning of reactors. The study places in perspective the size of the job, the hazards involved, the cost and the environmental impact. The three internationally agreed ''stages'' of decommissioning, namely, mothballing, entombment, and dismantling are defined and discussed. The single unit 600 MW(e) CANDU is chosen as the type of reactor on which the discussion is focussed but the conclusions reached will provide a basis for judgement of the costs and problems associated with decommissioning reactors of other sizes and types. (author)

Measurements of low reactivities using a reactor oscillator

International Nuclear Information System (INIS)

Obradovic, D.; Petrovic, M.

1965-12-01

Most of the methods of measuring reactivity are limited to the region from several hundreds to several thousands of pcm. The present work develops a method of measuring low reactivities from several pcm to about 600 pcm using the ROB-1 reactor oscillator on the RB reactor of the Boris Kidric Institute of Nuclear Sciences at Vinca. The accuracy of measurement is better than 1%. Several methods are used to measure low reactivities. The most often used is the method based on measuring the stable reactor period. The bottom limit of this method is about 30 porn /1,2/. For control rod calibration the method of rod oscillation is used /3,4/. This method is confronted with considerable influence of space effects /5/. Reference /6/ reports on a method for measuring the reactivity coefficient at a critical level in liquid-moderated reactors. The method is based on measuring reactor response to the oscillation of the moderator about the critical level. The present work reports on a method of determining the reactivity by measuring the phase shift between the perturbation of the effective multiplication factor and reactor response. With the use of the ROB-1 reactor oscillator, the method allows measurement of the reactivity from several pcm to about 600 pcm with an accuracy of 1% (author)

Measurements of low reactivities using a reactor oscillator

Energy Technology Data Exchange (ETDEWEB)

Obradovic, D; Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-12-15

Most of the methods of measuring reactivity are limited to the region from several hundreds to several thousands of pcm. The present work develops a method of measuring low reactivities from several pcm to about 600 pcm using the ROB-1 reactor oscillator on the RB reactor of the Boris Kidric Institute of Nuclear Sciences at Vinca. The accuracy of measurement is better than 1%. Several methods are used to measure low reactivities. The most often used is the method based on measuring the stable reactor period. The bottom limit of this method is about 30 porn /1,2/. For control rod calibration the method of rod oscillation is used /3,4/. This method is confronted with considerable influence of space effects /5/. Reference /6/ reports on a method for measuring the reactivity coefficient at a critical level in liquid-moderated reactors. The method is based on measuring reactor response to the oscillation of the moderator about the critical level. The present work reports on a method of determining the reactivity by measuring the phase shift between the perturbation of the effective multiplication factor and reactor response. With the use of the ROB-1 reactor oscillator, the method allows measurement of the reactivity from several pcm to about 600 pcm with an accuracy of 1% (author)

Quality of Life in Patients With Brain Metastases Using the EORTC QLQ-BN20+2 and QLQ-C15-PAL

International Nuclear Information System (INIS)

Caissie, Amanda; Nguyen, Janet; Chen, Emily; Zhang Liying; Sahgal, Arjun; Clemons, Mark; Kerba, Marc; Arnalot, Palmira Foro; Danjoux, Cyril; Tsao, May; Barnes, Elizabeth; Holden, Lori; Danielson, Brita; Chow, Edward

2012-01-01

Purpose: The 20-item European Organisation for Research and Treatment of Cancer Quality of Life Questionnaire–Brain Neoplasm (QLQ-BN20) is a validated quality-of-life (QOL) questionnaire for patients with primary brain tumors. The European Organisation for Research and Treatment of Cancer Quality of Life Questionnaire–Core 15 Palliative (QLQ-C15-PAL) core palliative questionnaire is a 15-item version of the core 30-item QLQ-C30 and was developed to decrease the burden on patients with advanced cancer. The combination of the QLQ-BN20 and QLQ-C30 to assess QOL may be too burdensome for patients. The primary aim of this study was to assess QOL in patients before and after treatment for brain metastases using the QLQ-BN20+2 and QLQ-C15-PAL, a version of the QLQ-BN20 questionnaire with 2 additional questions assessing cognitive functioning that were not addressed in the QLQ-C15-PAL. Methods and Materials: Patients with brain metastases completed the QLQ-C15-PAL and QLQ-BN20+2 questionnaires to assess QOL before and 1 month after radiation. Linear regression analysis was used to assess changes in QOL scores over time, as well as to explore associations between the QLQ-BN20+2 and QLQ-C15-PAL scales, patient demographics, and clinical variables. Spearman correlation assessed associations between the QLQ-BN20+2 and QLQ-C15-PAL scales. Results: Among 108 patients, the majority (55%) received whole-brain radiotherapy only, with 65% of patients completing follow-up at 1 month after treatment. The most prominent symptoms at baseline were future uncertainty (QLQ-BN20+2) and fatigue (QLQ-C15-PAL). After treatment, significant improvement was seen for the QLQ-C15-PAL insomnia scale, as well as the QLQ-BN20+2 scales of future uncertainty, visual disorder, and concentration difficulty. Baseline Karnofsky Performance Status was negatively correlated to QLQ-BN20+2 motor dysfunction but positively related to QLQ-C15-PAL physical functioning and QLQ-BN20+2 cognitive functioning at

Quality of Life in Patients With Brain Metastases Using the EORTC QLQ-BN20+2 and QLQ-C15-PAL

Energy Technology Data Exchange (ETDEWEB)

Caissie, Amanda; Nguyen, Janet; Chen, Emily; Zhang Liying [Rapid Response Radiotherapy Program, Odette Cancer Centre, Sunnybrook Health Sciences Centre, University of Toronto, Toronto, Ontario (Canada); Sahgal, Arjun [Rapid Response Radiotherapy Program, Odette Cancer Centre, Sunnybrook Health Sciences Centre, University of Toronto, Toronto, Ontario (Canada); Department of Radiation Oncology, Princess Margaret Hospital, University of Toronto, Toronto, Ontario (Canada); Clemons, Mark [Department of Medical Oncology, Ottawa Hospital Cancer Centre, Ottawa, Ontario (Canada); Kerba, Marc [Department of Radiation Oncology, Tom Baker Cancer Centre, Calgary, Alberta (Canada); Arnalot, Palmira Foro [Parc de Salut Mar Hospital de l' Esperanca, Barcelona (Spain); Danjoux, Cyril; Tsao, May; Barnes, Elizabeth; Holden, Lori [Rapid Response Radiotherapy Program, Odette Cancer Centre, Sunnybrook Health Sciences Centre, University of Toronto, Toronto, Ontario (Canada); Danielson, Brita [Department of Radiation Oncology, Cross Cancer Institute, Edmonton, Alberta (Canada); Chow, Edward, E-mail: edward.chow@sunnybrook.ca [Rapid Response Radiotherapy Program, Odette Cancer Centre, Sunnybrook Health Sciences Centre, University of Toronto, Toronto, Ontario (Canada)

2012-07-15

Purpose: The 20-item European Organisation for Research and Treatment of Cancer Quality of Life Questionnaire-Brain Neoplasm (QLQ-BN20) is a validated quality-of-life (QOL) questionnaire for patients with primary brain tumors. The European Organisation for Research and Treatment of Cancer Quality of Life Questionnaire-Core 15 Palliative (QLQ-C15-PAL) core palliative questionnaire is a 15-item version of the core 30-item QLQ-C30 and was developed to decrease the burden on patients with advanced cancer. The combination of the QLQ-BN20 and QLQ-C30 to assess QOL may be too burdensome for patients. The primary aim of this study was to assess QOL in patients before and after treatment for brain metastases using the QLQ-BN20+2 and QLQ-C15-PAL, a version of the QLQ-BN20 questionnaire with 2 additional questions assessing cognitive functioning that were not addressed in the QLQ-C15-PAL. Methods and Materials: Patients with brain metastases completed the QLQ-C15-PAL and QLQ-BN20+2 questionnaires to assess QOL before and 1 month after radiation. Linear regression analysis was used to assess changes in QOL scores over time, as well as to explore associations between the QLQ-BN20+2 and QLQ-C15-PAL scales, patient demographics, and clinical variables. Spearman correlation assessed associations between the QLQ-BN20+2 and QLQ-C15-PAL scales. Results: Among 108 patients, the majority (55%) received whole-brain radiotherapy only, with 65% of patients completing follow-up at 1 month after treatment. The most prominent symptoms at baseline were future uncertainty (QLQ-BN20+2) and fatigue (QLQ-C15-PAL). After treatment, significant improvement was seen for the QLQ-C15-PAL insomnia scale, as well as the QLQ-BN20+2 scales of future uncertainty, visual disorder, and concentration difficulty. Baseline Karnofsky Performance Status was negatively correlated to QLQ-BN20+2 motor dysfunction but positively related to QLQ-C15-PAL physical functioning and QLQ-BN20+2 cognitive functioning at

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

International Nuclear Information System (INIS)

Mayorshin, A.A.; Kisly, V.A.; Sudakov, L.V.

2000-01-01

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)

Design of High Temperature Reactor Vessel Using ANSYS Software

International Nuclear Information System (INIS)

Bandriyana; Kasmudin

2003-01-01

Design calculation and evaluation of material strength for high temperature reactor vessel based on the design of HTR-10 high temperature reactor vessel were carried out by using the ANSYS 5.4 software. ANSYS software was applied to calculate the combined load from thermal and pressure load. Evaluation of material strength was performed by calculate and determine the distribution of temperature, stress and strain in the thickness direction of vessel, and compared with its material strength for designed. The calculation was based on the inner wall temperature of vessel of 600 o C and the outer temperature of 500 and 600 o C. Result of calculation gave the maximum stress for outer temperature of 600 o C was 288 N/ mm 2 and strain of 0.000187. For outer temperature of 500 o C the maximum stress was 576 N/ mm 2 and strain of 0.003. Based on the analysis result, the material of steel SA 516-70 with limited stress for design of 308 N/ mm 2 can be used for vessel material with outer wall temperature of 600 o C

Time-resolved Polarimetry of the Superluminous SN 2015bn with the Nordic Optical Telescope

DEFF Research Database (Denmark)

Leloudas, Giorgos; Maund, Justyn R.; Gal-Yam, Avishay

2017-01-01

We present imaging polarimetry of the superluminous supernova SN 2015bn, obtained over nine epochs between -20 and +46 days with the Nordic Optical Telescope. This was a nearby, slowly evolving Type I superluminous supernova that has been studied extensively and for which two epochs of spectropol......We present imaging polarimetry of the superluminous supernova SN 2015bn, obtained over nine epochs between -20 and +46 days with the Nordic Optical Telescope. This was a nearby, slowly evolving Type I superluminous supernova that has been studied extensively and for which two epochs...... of spectropolarimetry are also available. Based on field stars, we determine the interstellar polarization in the Galaxy to be negligible. The polarization of SN 2015bn shows a statistically significant increase during the last epochs, confirming previous findings. Our well-sampled imaging polarimetry series allows us...

Security-by-design approach of the KALIMER 600 SFR plant

International Nuclear Information System (INIS)

So, Dong Sup; Lee, Yong Bum

2012-01-01

Security measures as well as safety and safeguards measures should be incorporated and addressed early in the design process to enhance the cost effectiveness of a PPS (Physical Protection System). Safety, security, operations, and safeguards design teams and regulators need to be flexible and perform 'trade studies' on the available options. In this paper, SBD (Security by Design) measures in the design phase of the KALIMER 600 SFR (Sodium Cooled Reactor) plant are identified and discussed qualitatively

Novel composite cBN-TiN coating deposition method: structure and performance in metal cutting

International Nuclear Information System (INIS)

Russell, W.C.; Malshe, A.P.; Yedave, S.N.; Brown, W.D.

2001-01-01

Cubic boron nitride coatings are under development for a variety of applications but stabilization of the pure cBN form and adhesion of films deposited by PVD and ion-based methods has been difficult. An alternative method for depositing a composite cBN-TiN film has been developed for wear related applications. The coating is deposited in a two-stage process utilizing ESC (electrostatic spray coating) and CVI (chemical vapor infiltration). Fully dense films of cBN particles evenly dispersed in a continuous TiN matrix have been developed. Testing in metal cutting has shown an increase in tool life (turning - 4340 steel) of three to seven times, depending of machining parameters, in comparison with CVD deposited TiN films. (author)

Monte-Carlo modeling of parameters of a subcritical cascade reactor based on MSBR and LMFBR technologies

International Nuclear Information System (INIS)

Bznuni, S.A.; Zhamkochyan, V.M.; Khudaverdyan, A.G.; Barashenkov, V.S.; Sosnin, A.N.; Polanski, A.

2001-01-01

Parameters are investigated of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k eff = 0.94 - 0.98), is capable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10 14 cm 12 · s -1 , in the fast booster zone is 5.12 · 10 15 cm 12 · s -1 at k eff = 0.98 and proton beam current I = 2.1 mA. (author)

Monte-Carlo Modeling of Parameters of a Subcritical Cascade Reactor Based on MSBR and LMFBR Technologies

CERN Document Server

Bznuni, S A; Zhamkochyan, V M; Polanski, A; Sosnin, A N; Khudaverdyan, A H

2001-01-01

Parameters of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k_{eff}=0.94-0.98), is apable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10^{14} cm^{12}\\cdot s^_{-1}, in the fast booster zone is 5.12\\cdot10^{15} cm^{12}\\cdot s{-1} at k_{eff}=0.98 and proton beam current I=2.1 mA.

Further delays hit troubled $2bn cosmic-ray detector

CERN Multimedia

Cartlidge, Edwin

2010-01-01

"A $2bn mission to study cosmic rays will have to wait another few months before being sent to the International Space Station (ISS) after NASA announced last month that it was pushing back the launch of the Space Shuttle Endeavour until 26 February 2011" (0.5 page)

Stabilization of the O p2x2 phase on Cu(001) sheltered by wrinkled BN over-layer

Science.gov (United States)

Kim, Yong-Sung; Ma, Chuanxu; Li, An-Ping; Yoon, Mina

The 2 √3x √3R45°phase of oxygen (O) on the Cu(001) surface has been observed in scanning tunneling microscopy (STM) measurements. Although the p2x2 phase of O on the Cu(001) surface has been proposed theoretically to be the most stable in O-lean conditions, it has not been observed in experiments for a long time. Recently, the O p2x2 phase has been found in STM on the Cu(001) surface with an overlying BN monolayer. In this theoretical study, we investigate what the role of BN over-layer is to stabilize the O p2x2 phase on the Cu(001) surface. The BN over-layer is lattice-matched with the Cu(001) surface and the BN mono-layer sheet is periodically wrinkled along the BN arm-chair direction and along the [100] or [010] direction on the Cu(001) surface. The interlayer space between the Cu(001) surface and the bulge of the wrinkled BN sheet is found to play as a preferential shelter for O to be adsorbed, and the boundary of the BN inner wall along the [010] or [100] direction makes the p2x2 phase more favorable against the 45°-tilted 2 √3x √3R45°phase of O on the Cu(001) surface. This was supported by Center for Nanophase Materials Sciences, which is a DOE Office of Science User Facility, and the Laboratory Directed Research and Development Program of Oak Ridge National Laboratory, maaged by UT-Battelle, LLC, for the U. S. DOE.

Parametric Study on an Initial Cooling Performance in the KALIMER-600

International Nuclear Information System (INIS)

Han, Ji-Woong; Eoh, Jae-Hyuk; Lee, Tae-Ho; Kim, Seong-O

2009-01-01

Decay heat removal is very important in a nuclear power plant. The KALIMER-600, Korea Advanced Liquid MEtal Reactor, employs the PDRC(Passive Decay heat Removal Circuit) to remove the decay heat. DHX(Decay Heat eXchanger) in the PDRC of KALIMER-600 is disposed in the DHX support barrel located in the hot pool region. Each DHX support barrel has the lower end communicating with the cold pool such that the sodium free surface inside the barrel is maintained with the same level of the cold pool using the pumping head of the PHTS(Primary Heat Transport System) pumps. Consequently, DHX is not in direct contact with the cold pool sodium during a normal plant operation. Under transient conditions such as the loss of a normal heat sink accident, free surface outside the barrel rises up due to the expansion of the sodium induced by the core decay heat during the initial stage cooling. When it overflows into the cold pool through the DHX support barrel the heat removal via DHX is initiated and the second stage cooling begins. In order to secure the safety of a reactor until the activation of a second stage cooling by PDRC, it is very important to suppress the core temperature rising by an enhancement of the initial cooling performance. In this study the parametric investigations have been applied to reveal the effect of various design parameters on the initial cooling performance. The various design parameters such as coastdown flow, IHX(Intermediate Heat eXchanger) elevation, heat transfer via CCS (Cavity Cooling System) were considered. The numerical approaches based on a multidimensional analysis can be utilized as a useful tool to investigate overall transient behaviors within a pool. In this research the COMMIX-1AR/P code is utilized as a transient analysis tool in KALIMER-600 after a shut down. This study will provide the basic design information to improve the initial cooling performance in the KALIMER-600

Studies for the layout and technical conception of a two-circuit HTR power plant of 600 MWsub(el) under public utilizer aspects

International Nuclear Information System (INIS)

Schuetten, R.

1981-01-01

In this study concerning conceptions for a nuclear power plant of 600 MWsub(el) with high-temperature reactor a conception for a HTR-nuclear power plant of 600 MWsub(el) to be built in the Federal Republic of Germany in future is developed on the basis of operating experience with the 15-MW-AVR-experimental nuclear power plant, the construction of the THTR-300 nuclear power plant and the gas-cooled reactors of English, French and American origin. This report gives a survey of the most important findings and the requirements on behalf of the public utilities for a nuclear power plant with high-temperature reactor with the dimensions of 600 MWsub(el). The examination of the utilities basic requirements for a power plant and the experience made during the licensing procedure led to this technical and safety conception for a HTR nuclear power plant with spherical fuel elements. In addition, the questions of the possibility of recurrent tests and of repairing safety components and also the future shut-down of the power plant, which are important for the public utilities, are examined. (orig./GL) [de

MINIMARS: An attractive small tandem mirror fusion reactor

International Nuclear Information System (INIS)

Perkins, L.J.; Logan, B.G.; Doggett, J.N.; Devoto, R.S.

1986-01-01

Through the innovative design of a novel end plug scheme employing octopole MHD stabilization, the authors present the conceptual design of ''MINIMARS'', a small commercial fusion reactor based on the tandem mirror principle. The current baseline for MINIMARS has a net electric output of 600 MWe and they have configured the design for short construction times, factory-built modules, inherently safe blanket systems, and multiplexing in station sizes of ≅ 600-2400 MWe. They demonstrate that the compact octopole end cell provides a number of advantages over the more conventional quadrupole (yin-yang) end cell encountered in the MARS tandem mirror reactor study, and enables ignition to be achieved with much shorter central cell lengths. Accordingly, being economic in small sizes, MINIMARS provides an attractive alternative to the more conventional larger conceptual fusion reactors encountered to date, and would contribute significantly to the lowering of utility financial risk in a developing fusion economy

First-Principles Investigations of the Working Mechanism of 2D h-BN as an Interfacial Layer for the Anode of Lithium Metal Batteries.

Science.gov (United States)

Shi, Le; Xu, Ao; Zhao, Tianshou

2017-01-18

An issue with the use of metallic lithium as an anode material for lithium-based batteries is dendrite growth, causing a periodic breaking and repair of the solid electrolyte interphase (SEI) layer. Adding 2D atomic crystals, such as h-BN, as an interfacial layer between the lithium metal anode and liquid electrolyte has been demonstrated to be effective to mitigate dendrite growth, thereby enhancing the Columbic efficiency of lithium metal batteries. But the underlying mechanism leading to the reduced dendrite growth remains unknown. In this work, with the aid of first-principle calculations, we find that the interaction between the h-BN and lithium metal layers is a weak van der Waals force, and two atomic layers of h-BN are thick enough to block the electron tunneling from lithium metal to electrolyte, thus prohibiting the decomposition of electrolyte. The interlayer spacing between the h-BN and lithium metal layers can provide larger adsorption energies toward lithium atoms than that provided by bare lithium or h-BN, making lithium atoms prefer to intercalate under the cover of h-BN during the plating process. The combined high stiffness of h-BN and the low diffusion energy barriers of lithium at the Li/h-BN interfaces induce a uniform distribution of lithium under h-BN, therefore effectively suppressing dendrite growth.

Quasi-Two-Dimensional h-BN/β-Ga2O3 Heterostructure Metal-Insulator-Semiconductor Field-Effect Transistor.

Science.gov (United States)

Kim, Janghyuk; Mastro, Michael A; Tadjer, Marko J; Kim, Jihyun

2017-06-28

β-gallium oxide (β-Ga 2 O 3 ) and hexagonal boron nitride (h-BN) heterostructure-based quasi-two-dimensional metal-insulator-semiconductor field-effect transistors (MISFETs) were demonstrated by integrating mechanical exfoliation of (quasi)-two-dimensional materials with a dry transfer process, wherein nanothin flakes of β-Ga 2 O 3 and h-BN were utilized as the channel and gate dielectric, respectively, of the MISFET. The h-BN dielectric, which has an extraordinarily flat and clean surface, provides a minimal density of charged impurities on the interface between β-Ga 2 O 3 and h-BN, resulting in superior device performances (maximum transconductance, on/off ratio, subthreshold swing, and threshold voltage) compared to those of the conventional back-gated configurations. Also, double-gating of the fabricated device was demonstrated by biasing both top and bottom gates, achieving the modulation of the threshold voltage. This heterostructured wide-band-gap nanodevice shows a new route toward stable and high-power nanoelectronic devices.

Spin polarization of graphene and h -BN on Co(0001) and Ni(111) observed by spin-polarized surface positronium spectroscopy

Science.gov (United States)

Miyashita, A.; Maekawa, M.; Wada, K.; Kawasuso, A.; Watanabe, T.; Entani, S.; Sakai, S.

2018-05-01

In spin-polarized surface positronium annihilation measurements, the spin polarizations of graphene and h -BN on Co(0001) were higher than those on Ni(111), while no significant differences were seen between graphene and h -BN on the same metal. The obtained spin polarizations agreed with those expected from first-principles calculations considering the positron wave function and the electron density of states from the first surface layer to the vacuum region. The higher spin polarizations of graphene and h -BN on Co(0001) as compared to Ni(111) simply reflect the spin polarizations of these metals. The comparable spin polarizations of graphene and h -BN on the same metal are attributed to the creation of similar electronic states due to the strong influence of the metals: the Dirac cone of graphene and the band gap of h -BN disappear as a consequence of d -π hybridization.

Carbon-tuned bonding method significantly enhanced the hydrogen storage of BN-Li complexes.

Science.gov (United States)

Deng, Qing-ming; Zhao, Lina; Luo, You-hua; Zhang, Meng; Zhao, Li-xia; Zhao, Yuliang

2011-11-01

Through first-principles calculations, we found doping carbon atoms onto BN monolayers (BNC) could significantly strengthen the Li bond on this material. Unlike the weak bond strength between Li atoms and the pristine BN layer, it is observed that Li atoms are strongly hybridized and donate their electrons to the doped substrate, which is responsible for the enhanced binding energy. Li adsorbed on the BNC layer can serve as a high-capacity hydrogen storage medium, without forming clusters, which can be recycled at room temperature. Eight polarized H(2) molecules are attached to two Li atoms with an optimal binding energy of 0.16-0.28 eV/H(2), which results from the electrostatic interaction of the polarized charge of hydrogen molecules with the electric field induced by positive Li atoms. This practical carbon-tuned BN-Li complex can work as a very high-capacity hydrogen storage medium with a gravimetric density of hydrogen of 12.2 wt%, which is much higher than the gravimetric goal of 5.5 wt % hydrogen set by the U.S. Department of Energy for 2015.

Study on core make-up water experiment of AC600 make-up water tank

International Nuclear Information System (INIS)

Ji Fuyun; Li Changlin; Zheng Hua; Liu Shaohua; Xu Xiaolan

1999-01-01

The core makeup tank (CMT) is a principal component of the passive high pressure safety injection systems for AC600 and has a function to inject cold borated water into reactor vessel during abnormal events. The purpose of this experiment is to verify the gravity drain behavior of the CMT and to provide experimental data to verify the computer codes used in the safety analyses. Five experiments with simulative small and medium break conditions are conducted at AC600 core makeup tank performance test facility of Nuclear Power Institute of China (NPIC). The author provides the results of one test. The simulated accident is a small break loss-of-coolant accident

Thermosyphoning in the CANDU reactor

International Nuclear Information System (INIS)

Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

1984-01-01

Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

Workshop UNK-600 (proceedings); Materialy rabochego soveshchaniya UNK-600

Energy Technology Data Exchange (ETDEWEB)

Zajtsev, A M; Bitykov, S I [eds.

1994-12-31

Proceedings are presented of the workshop devoted to the accelerating storage complex of IHEP (UNK-600). In the first section is given the information on the present status of the UNK-600 and particle channels design and on the adopted experiment NEPTUN-A. In the papers of the second section are discussed hadron physics investigations at 600 GeV. Experiments in the neutrino and muon beams are analyzed. A possible program of studying the charged kaon rare decays is described.

Fabrication and characterization of Ni-decorated h-BN powders with ChCl-EG ionic liquid as addition by electroless deposition

Science.gov (United States)

Yang, Qionglian; Ru, Juanjian; Song, Peng; Hu, Mingyu; Feng, Jing

2018-05-01

Ni-decorated h-BN powders are fabricated with ChCl-EG as additive via electroless plating in the paper. As comparison, the different additive concentration of choline chloride-ethylene glycol (ChCl-EG) ionic liquid (0 g l-1, 30 g l-1, 60 g l-1, 90 g l-1) is presented. The effects of ChCl-EG concentration are studied, including the surface morphologies, phase analysis of Ni-decorated h-BN powders and the residual Ni2+ concentration is measured in electroless plating bath. It is demonstrated that the deposition phenomena of nickel particles on h-BN surface is changed with the addition of ChCl-EG. When the concentration of ChCl-EG is 30 g l-1, the Ni particles on h-BN surface are in dispersed and spheroid state with the average size of 10-1000 nm. It can be found that 30 g l-1 ChCl-EG is conducive to the arise of deposition phenomena, which is the formation of the single nickel particle on h-BN surface. Besides, more Ni particles are deposited on h-BN surface with the increase of nickel plating times, which is characterized with scanning electron microscope and transmission electron microscope. Furthermore, the deposition phenomenon and growth mechanism are proposed without and with ChCl-EG as additive to further elaborate the formation of Ni particles on h-BN surface.

Fabrication and characterization of Ni-decorated h-BN powders with ChCl-EG ionic liquid as addition by electroless deposition.

Science.gov (United States)

Yang, Qionglian; Ru, Juanjian; Song, Peng; Hu, Mingyu; Feng, Jing

2018-05-01

Ni-decorated h-BN powders are fabricated with ChCl-EG as additive via electroless plating in the paper. As comparison, the different additive concentration of choline chloride-ethylene glycol (ChCl-EG) ionic liquid (0 g l -1 , 30 g l -1 , 60 g l -1 , 90 g l -1 ) is presented. The effects of ChCl-EG concentration are studied, including the surface morphologies, phase analysis of Ni-decorated h-BN powders and the residual Ni 2+ concentration is measured in electroless plating bath. It is demonstrated that the deposition phenomena of nickel particles on h-BN surface is changed with the addition of ChCl-EG. When the concentration of ChCl-EG is 30 g l -1 , the Ni particles on h-BN surface are in dispersed and spheroid state with the average size of 10-1000 nm. It can be found that 30 g l -1 ChCl-EG is conducive to the arise of deposition phenomena, which is the formation of the single nickel particle on h-BN surface. Besides, more Ni particles are deposited on h-BN surface with the increase of nickel plating times, which is characterized with scanning electron microscope and transmission electron microscope. Furthermore, the deposition phenomenon and growth mechanism are proposed without and with ChCl-EG as additive to further elaborate the formation of Ni particles on h-BN surface.

Results and Prospects of Development of Works on Structural Core Materials for Russian Fast Reactors

International Nuclear Information System (INIS)

Nikitina, A.A.; Ageev, V.S.; Leontyeva-Smirnova, M.V.; Mitrofanova, N.M.; Tselishchev, A.V.

2015-01-01

The strategy of development of atomic energy in Russia in the first half of XXI century contemplates construction and putting in operation of fast reactors of new generation with different types of coolant: sodium (BN-800, BN-1200, MBIR), lead (BREST-OD-300) and lead-bismuth eutectic (SVBR-100). For assurance of the working capacity of reactors that are under construction and achievement of economically reasonable burn-up of nuclear fuel the structural core materials with necessary level of radiation resistance, heat resistance, corrosion resistance to products of fuel fission, corrosion resistance in coolant and in water must be developed and justified. For sodium cooled reactors the key challenge is creation of radiation resistant and heat resistant cladding materials, which must ensure the achievement of damage doses at least 140 dpa. The solution of this problem is provided by phased use as cladding materials of austenitic steels ChS68 and EK164 (maximum damage doses ~ 92 and ~110-115 dpa, respectively), precipitation-hardening heat resistant ferritic-martensitic steels EK181 and ChS139 (maximum damage dose ~140 dpa) and oxide dispersion strengthened (ODS) steels (maximum damage dose more than 140 dpa). For development of core materials for reactors with lead and lead-bismuth eutectic coolants the most serious challenge is corrosion resistance of materials in coolant. Therefore at present time a very wide range of works on study of corrosion resistance of candidate materials is carrying out. As the basic material for the cladding tubes is considered a ferritic-martensitic steel EP823 with high silicon content. In this report the main results of works on justification of the working capacity of materials of different classes in respect to use it in cores of operating and prospective fast reactors with different types of coolant and prospects of further development of works are presented. (author)

Conservative Analysis of TOP and LOF for KALIMER-600 with the SSC-K code

International Nuclear Information System (INIS)

Jeong, H. Y.; Ha, K. S.; Kwon, Y. M.; Suk, S. D.; Lee, K. L.; Lee, Y. B.; Cho, C. H.

2009-01-01

KALIMER-600 is designed to satisfy the safety principle of a defense-in-depth and also the safety design objectives which have been established to implement the safety principle in the design. Highly reliable diversified shutdown mechanisms are equipped for the reactivity control function during an accident or abnormal transients in KALIMER-600. The reactivity is also controlled by the inherent reactivity feedback mechanisms incorporated in the design. In addition, a uniquely designed passive decay heat removal circuit provides the heat removal function. Due to these passive and inherent safety characteristics, the safety of KALIMER-600 is much improved than the existing PWR designs. Therefore, the events whose frequencies are higher than 10 -7 per reactor-year are categorized as design basis events (DBEs). The safety analysis has been performed for the TOP and LOF events which are two most important DBEs in KALIMER-600. The analysis results show that the fuel, clad, and the coolant temperatures are well within the safety limit temperatures. Therefore, the KALIMER-600 design fulfills the design basis safety criteria with no fuel damage and no threat to its structural integrity during the transients. Through the analysis, it is clearly shown that the KALIMER-600 design maintains its safety functions required for the mitigation of accidents with an appropriate margin. Therefore, it is concluded that the KALIMER-600 breakeven core design ensures the safety margins for the considered DBEs

Increasing the reliability, availability, and maintainability of the AP600 by design

International Nuclear Information System (INIS)

Trombola, D.; Meyer, C.

1993-01-01

The AP600 design is based on providing a safe, simple, standardized, and economically competitive design with a high degree of operability and ease of maintenance. Design features such as component selection, layout, and standardization increase the probability that targeted repair times are achieved. Design requirements from the utility industry and industry design practices have established criteria for: layout, changeout and replacement of parts and components; access for major pieces of equipment; and vehicle passage. These features coupled with a solid reliability assurance and maintenance program will help the AP600 meet its objectives for operation and maintenance. The AP600 draws on the operating experience and lessons learned from the utility community through design workshops and design review interaction, as well as operating plant data from sources several sources. Internally, the AP600 program incorporates the resources of Westinghouse NSD (Nuclear Service Division), which for decades has provided refueling, steam generator, reactor coolant pump, and other operating plant services. Since the early phases of the design process, the AP600 Program has executed a comprehensive reliability, availability, and maintainability program (RAM) which dealt primarily with assessing and improving plant availability. In conjunction with this program a Probabilistic Risk Assessment (PRA) was performed and submitted to the NRC with the Standard Safety Analysis Report (SSAR) in June 1992. This paper describes how AP600 ensures that the plant has design features to enhance reliability, availability, and maintainability. The RAM program that brings the plant through the design certification phase is described

Dielectric Response and Born Dynamic Charge of BN Nanotubes from Ab Initio Finite Electric Field Calculations

Science.gov (United States)

Guo, Guang-Yu; Ishibashi, Shoji; Tamura, Tomoyuki; Terakura, Kiyoyuki

2007-03-01

Since the discovery of carbon nanotubes (CNTs) in 1991 by Iijima, carbon and other nanotubes have attracted considerable interest worldwide because of their unusual properties and also great potentials for technological applications. Though CNTs continue to attract great interest, other nanotubes such as BN nanotubes (BN-NTs) may offer different opportunities that CNTs cannot provide. In this contribution, we present the results of our recent systematic ab initio calculations of the static dielectric constant, electric polarizability, Born dynamical charge, electrostriction coefficient and piezoelectric constant of BN-NTs using the latest crystalline finite electric field theory [1]. [1] I. Souza, J. Iniguez, and D. Vanderbilt, Phys. Rev. Lett. 89, 117602 (2002); P. Umari and A. Pasquarello, Phys. Rev. Lett. 89, 157602 (2002).

Travelling cranes for heavy reactor component handling

International Nuclear Information System (INIS)

Champeil, M.

1977-01-01

Structure and operating machinery of two travelling cranes (600 t and 450 t) used in the Framatome factory for handling heavy reactor components are described. When coupled, these cranes can lift loads up to 1000 t [fr

Two important safety-related verification tests in the design of Qinshan NPP 600 MWe reactor

International Nuclear Information System (INIS)

Li Pengzhou; Li Tianyong; Yu Danping; Sun Lei

2005-01-01

This paper summarizes two most important verification tests performed in the design of reactor of Qinshan NPP Phase II: seismic qualification test of control rod drive line (CRDL), flow-induced vibration test of reactor internals both in 1:5 scaled model and on-site measurement during heat function testing (HFT). Both qualification tests proved that the structural design of the reactor has large safety margin. (authors)

Thermal Aging Effects on Residual Stress and Residual Strain Distribution on Heat Affected Zone of Alloy 600 in Dissimilar Metal Weld

Energy Technology Data Exchange (ETDEWEB)

Ham, Junhyuk; Choi, Kyoung Joon; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-10-15

Dissimilar metal weld (DMW), consisting of Alloy 600, Alloy 182, and A508 Gr.3, has been widely used as a joining material of the reactor pressure vessel penetration nozzle and the steam generator tubing for pressurized water reactors (PWR) because of its good mechanical strength, thermal conductivity, and corrosion resistance. Residual tensile stress is mainly nominated as a cause of SCC in light water reactors by IAEA report. So, to relax the residual stress, post-weld heat treatment is required after manufacturing process such as welding. However, thermal treatment has a great effect on the microstructure and the chromium depletion profile on Alloy 600, so called sensitization. By this reason, HAZ on Alloy 600 is critical to crack. According to G.A. Young et al., Crack growth rates (CGR) in the Alloy 600 HAZ were about 30 times faster than those in the Alloy 600 base metal tested under the same conditions. And according to Z.P. Lu et al., CGR in the Alloy 600 HAZ can be more than 20 times higher than that in its base metal. There are some methods to measure the exact value of residual stress on the material surface. The most common way is X-ray diffraction method (XRD). The principle of XRD is based on lattice strains and depends on the changes in the spacing of the atomic planes in material. And there is a computer simulation method to estimate residual stress distribution which is called ANSYS. This study was conducted to investigate how thermal aging affects residual stress and residual strain distribution of Alloy 600 HAZ. Following conclusions can be drawn from this study. According to preceding researches and this study, both the relaxation of residual stress and the change of residual strain follow as similar way, spreading out from concentrated region. The result of Vickers micro-hardness tester shows that tensile residual stresses are distributed broadly on the material aged by 15 years. Therefore, HT400{sub Y}15 material is weakest state for PWSCC. The

The influence of the water chemistry regime of the third circuit on the corrosion hydrogen burden to the secondary sodium circuit in the steam generator model of BN-800 reactor

International Nuclear Information System (INIS)

Smykov, V.B.; Ermolaev, N.P.; Kolesnik, A.I.; Egorov, V.A.; Shevchenko, N.N.

1994-01-01

An experimental program was conducted to determine the influence of water chemistry on the corrosion hydrogen burden from the III circuit to the secondary sodium in sodium-heated rig of OTSG of NPP BN-800. Combined water chemistry has given the best passivative effect on steam-generating surfaces and smallest hydrogen burden to secondary sodium during start-up. Common hydrogen increasing in secondary sodium was less then 0.2 ppm. In case of AVT water chemistry (NH 3 +N 2 H 4 ) in III side of OTSG-rig the hydrogen level in secondary sodium was 1.0-1.2 ppm. It means that during first start-up at NPP BN-800 the common hydrogen level in secondary sodium may reaches 0.80-0.85 ppm. 4 figs.; 4 tabs

Vickers Hardness of Diamond and cBN Single Crystals: AFM Approach

Directory of Open Access Journals (Sweden)

Sergey Dub

2017-12-01

Full Text Available Atomic force microscopy in different operation modes (topography, derivative topography, and phase contrast was used to obtain 3D images of Vickers indents on the surface of diamond and cBN single crystals with high spatial resolution. Confocal Raman spectroscopy and Kelvin probe force microscopy were used to study the structure of the material in the indents. It was found that Vickers indents in diamond has no sharp and clear borders. However, the phase contrast operation mode of the AFM reveals a new viscoelastic phase in the indent in diamond. Raman spectroscopy and Kelvin probe force microscopy revealed that the new phase in the indent is disordered graphite, which was formed due to the pressure-induced phase transformation in the diamond during the hardness test. The projected contact area of the graphite layer in the indent allows us to measure the Vickers hardness of type-Ib synthetic diamond. In contrast to diamond, very high plasticity was observed for 0.5 N load indents on the (001 cBN single crystal face. Radial and ring cracks were absent, the shape of the indents was close to a square, and there were linear details in the indent, which looked like slip lines. The Vickers hardness of the (111 synthetic diamond and (111 and (001 cBN single crystals were determined using the AFM images and with account for the elastic deformation of the diamond Vickers indenter during the tests.

Preliminary design concepts of an advanced integral reactor

International Nuclear Information System (INIS)

Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

1997-01-01

An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

Restrained shrinkage experiments on coated particle fuel compacts in the temperature range 600-1200 deg C

International Nuclear Information System (INIS)

Blackstone, R.; Veringa, H.J.; Loelgen, R.

1976-05-01

Information on irradiation induced creep in reactor graphite and in fuel compact material is an essential ingredient in the design of any reactor core layout, because the creep plasticity of these materials diminishes the stresses which are built up in the fuel element during reactor operation. The restrained shrinkage method in which the shrinkage of a dumbbell shaped creep specimen is restrained by a graphite material which shows less irradiation shrinkage, offers a good possibility of performing a large series of tensile creep experiments in a limited irradiation volume. The irradiations, evaluations and the results of a series of restrained shrinkage experiments in which six different materials were tested, of which five were dummy coated particle compacts and one pure matrix material are described and discussed. These materials were irradiated in the High Flux Reactor of the Euratom Joint Research Centre in Petten/Netherlands. The irradiations were performed in three successive capsules at irradiation temperatures of 600 deg C, 900 deg C, 1050 deg C and 1200 deg C up to a neutron fluence of maximum 3x10 21 n.cm 2 (DNE). The post-irradiation examinations yielded plastic strains up to 2,3%, and values for the radiation creep coefficient were calculated, ranging from 4 to 8.10 -12 at 600 deg C and 8 to 30.10 -12 at 1200 deg C always given per dyn.cm -2 tensile stresses and per 10 20 n.cm -2 fluence units. Generally it was found that the creep behavior of these materials and the temperature dependence of the creep process could be compared with those for normal reactor graphites

A review of fast reactor activities in Italy

International Nuclear Information System (INIS)

Tavoni, R.

1996-01-01

In this paper, Italian activities on liquid metal fast reactors are shown for the period May 1995 - April 1996. During this period the ENEA collaboration with General Electric on ALMR came to an end as a consequence of the reduced effort on the design development. Nevertheless ENEA completed the studies on the PRISM Mod B oxide burner core, the neutronic configuration of which was presented at last year's meeting. Some results of the dynamic calculations are shown. ENEA participated in the IAEA/EC benchmark on the comparative calculations for severe accident in BN-800 reactor. A complete neutronic evaluation has been made including power distribution, Doppler, sodium void and material coefficients. Activities on seismic isolation are also outlined. The Italian contribution to SPX restart and operation is described; some information about the complementary convention of the Nersa society is given, together with the Italian industry participation in the SPX restart. (author)

The biological characterization of 99mTc-BnAO-NI as a SPECT probe for imaging hypoxia in a sarcoma-bearing mouse model

International Nuclear Information System (INIS)

Hsia, Chien-Chung; Huang, Fu-Lei; Hung, Guang-Uei; Shen, Lie-Hang; Chen, Chuan-Lin; Wang, Hsin-Ell

2011-01-01

Objectives: Tumor growth beyond the region where vascular oxygen can reach creates a hypoxic domain. In this study, BnAO, a ligand that had been labeled with 99m Tc-pertechnetate for hypoxia imaging, was conjugated with 2-nitroimidazole to give 3,3,10,10-tetramethyl-1-(2-nitro-1H-imidazo-1-y1)-4,9-diazadodecane-2,11- dionedioxime (BnAO-NI) as a potential ligand for hypoxia detection. Pentoxifylline is a peripheral vasodilator and has been used as a radiosensitizer in tumor radiotherapy. 99m Tc-BnAO-NI/SPECT was applied to noninvasively assess the pharmacological effect of pentoxifylline in reducing tumor hypoxia in vivo. Methods: BnAO-NI was synthesized and formulated with methylene diphosphonate (MDP), stannous chloride and carbonate buffer to afford kits. After mixing with 99m Tc-pertechnetate, 99m Tc-BnAO-NI injection can be readily prepared. The partition coefficient, radiochemical purity and in vitro stability were determined. Cellular uptake of radiotracers in KHT cells under hypoxia was conducted in a CO 2 incubator at 37 o C under hypoxia or normoxia. A biodistribution study after intravenous injection of 99m Tc-BnAO-NI in KHT sarcoma-implanted C3H mice was performed. The effect of pentoxifylline (100 mg/kg) on reducing tumor hypoxia was also studied. Results: The radiochemical purity (RCP) of the 99m Tc-BnAO-NI preparation was greater than 96% and stable at ambient temperature for 24 h (RCP>90%). The accumulation of 99m Tc-BnAO-NI and 99m Tc-BnAO in KHT cells under hypoxia were 3.57 and 4.13-fold higher than those under normoxic environment, indicating unambiguous oxygen-dependent uptakes of these two probes. The distribution of 99m Tc-BnAO-NI in KHT sarcoma-bearing mice revealed rapid clearance from the blood circulation. The tumor uptake peaked at 2 h post-injection (0.32±0.05%ID/g) with tumor-to-blood and tumor-to-muscle ratios of 10.32 and 3.96, respectively. The effect of pentoxifylline on the tumor blood perfusion was obvious. The tumor