Research reactor spent nuclear fuel shipment from the Czech Republic to the Russian Federation

International Nuclear Information System (INIS)

Svoboda, K.; Broz, V.; Novosad, P.; Podlaha, J.; Svitak, F.

2009-01-01

In May 2004, the Global Threat Reduction Initiative agreement was signed by the governments of the United States and the Russian Federation. The goal of this initiative is to minimize, in cooperation with the International Atomic Energy Agency (IAEA) in Vienna, the existing threat of misuse of nuclear and radioactive materials for terrorist purposes, particularly highly enriched uranium (HEU), fresh and spent nuclear fuel (SNF), and plutonium, which have been stored in a number of countries. Within the framework of the initiative, HEU materials and SNF from research reactors of Russian origin will be transported back to the Russian Federation for reprocessing/liquidation. The program is designated as the Russian Research Reactor Fuel Return (RRRFR) Program and is similar to the U.S. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program, which is underway for nuclear materials of United States origin. These RRRFR activities are carried out under the responsibilities of the respective ministries (i.e., U.S. Department of Energy (DOE) and Russian Federation Rosatom). The Czech Republic and the Nuclear Research Institute Rez, plc (NRI) joined Global Threat Reduction Initiative in 2004. During NRI's more than 50 years of existence, radioactive and nuclear materials had accumulated and had been safely stored on its grounds. In 1995, the Czech regulatory body , State Office for Nuclear Safety (SONS), instructed NRI that all ecological burdens from its past activities must be addressed and that the SNF from the research reactor LVR -15 had to be transported for reprocessing. At the end of November 2007, all these activities culminated with the unique shipment to the Russian Federation of 527 fuel assemblies of SNF type EK-10 (enrichment 10% U-235) and IRT-M (enrichment 36% and 80% U-235) and 657 irradiated fuel rods of EK-10 fuel, which were used in LVR-15 reactor. (authors)

Calibration of spent fuel measurement assembly

International Nuclear Information System (INIS)

Koleska, Michal; Viererbl, Ladislav; Marek, Milan

2014-01-01

The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110m Ag isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110m Ag isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110m Ag isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

1997-01-01

The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

Study text and sets of questions for the training and examination of selected personnel of nuclear research facilities. Issue 2. Experimental teaching methods

International Nuclear Information System (INIS)

Fleischhans, J.; Hejzlar, R.; Hermansky, B.

2004-01-01

The VR-1 teaching reactor is described, 14 experimental exercises are given (e.g. Starting up and running the VR-1 reactor; Neutron detection and detectors; Measuring delayed neutrons ; Reactivity measurement; Control rod calibration; ...) and practical training at the existing Czech research reactors (LVR-15; LR-0; VR-1) is briefly highlighted. (P.A.)

Stability of the Hall sensors performance under neutron irradiation

International Nuclear Information System (INIS)

Duran, I.; Hron, M.; Stockel, J.; Viererbl, L.; Vsolak, R.; Cerva, V.; Bolshakova, I.; Holyaka, R.; Vayakis, G.

2004-01-01

A principally new diagnostic method must be developed for magnetic measurements in steady state regime of operation of fusion reactor. One of the options is the use of transducers based on Hall effect. The use of Hall sensors in ITER is presently limited by their questionable radiation and thermal stability. Issues of reliable operation in ITER like radiation and thermal environment are addressed in the paper. The results of irradiation tests of candidate Hall sensors in LVR-15 and IBR-2 experimental fission reactors are presented. Stable operation (deterioration of sensitivity below one percent) of the specially prepared sensors was demonstrated during irradiation by the total fluence of 3.10 16 n/cm 2 in IBR-2 reactor. Increasing the total neutron fluence up to 3.10 17 n/cm 2 resulted in deterioration of the best sensor's output still below 10% as demonstrated during irradiation in LVR-15 fission reactor. This level of neutron is already higher than the expected ITER life time neutron fluence for a sensor location just outside the ITER vessel. (authors)

Research activities at nuclear research institute in water chemistry and corrosion

International Nuclear Information System (INIS)

Kysela, Jan

2000-01-01

Research activities at Nuclear Research Institute Rez (NRI) are presented. They are based on former heavy water reactor program and now on pressurized reactors VVER types which are operated on Czech republic. There is LVR-15 research reactor operated in NRI. The reactor and its experimental facilities is utilized for water chemistry and corrosion studies. NRI services for power plants involve water chemistry optimalization, radioactivity build-up, fuel corrosion and structural materials corrosion tests. (author)

The BNCT project in the Czech Republic

International Nuclear Information System (INIS)

Burian, J.; Marek, M.; Rataj, J.; Honova, H.; Petruzelka, L.; Prokes, K.; Tovarys, F.; Dbaly, V.; Honzatko, J.; Tomandl, I.

2000-01-01

The start of clinical trials is expected before NCT Osaka 2000. The experiences from different part of project are presented. The BNCT facility at LVR-15 reactor of NRI consists of epithermal neutron beam with improved construction (6.98 x 10 8 /cm 2 s with acceptable background of fast neutrons and gammas) and irradiation and control rooms equipped by appropriate devices. Internationally-recognized software MacNCTPLAN is utilized for computational dosimetry and treatment planning. In the part of protocol the following parameters have been assessed: patient selection, BSH dosage, fractionation, starting dose, dose escalation steps. At the LVR-15, at horizontal channel, a prompt gamma ray analysis (PGRA) system has been developed and is operated for BNCT purposes. Some human blood samples were analyzed and compared with classical ICP method. During the process of licensing the experience was obtained, some notes are discussed in the paper. The first results were received for the study of biological effect of the LVR source for small animal model. (author)

Instruction texts and problems for the training and examination of selected personnel at research nuclear facilities

International Nuclear Information System (INIS)

Matejka, K.; Fleischhans, J.; Hejzlar, R.

1994-01-01

The publication comprises 6 separate brochures: (1) Selected chapters in reactor theory; (2) Experimental education methods; (3) Research and experimental reactors; (4.1) Technical description of the LVR-15 reactor; (4.2) Technical description of the LR-0 reactor; (4.3) Technical description of the VR-1 reactor; (5) Research reactor safety and operation; and (6) Database of problems for qualification examinations. Brochure No. 4 consists of 3 separate parts. The publication is intended for the training and examination of the following research reactor staff: reactor operator, shift engineer, control physicist, and start-up group head. (J.B.)

Dynamical model of computation of the rhodium self-powered neutron detector current

International Nuclear Information System (INIS)

Erben, O.; Slovacek, M.; Zerola, L.

1992-01-01

A model is presented for the calculation of the rhodium self-powered neutron detector current in dependence on the neutron flux density during reactor core transients. The total signal consists of a beta emission, prompt, and gamma component and a background signal. The model has been verified by means of experimental data obtained during measurements on the LVR-15 research reactor and at the Dukovany nuclear power plant. (author) 9 figs., 21 refs

Reduction in Left Ventricular Wall Stress and Improvement in Function in Failing Hearts using Algisyl-LVR

Science.gov (United States)

Lee, Lik Chuan; Zhihong, Zhang; Hinson, Andrew; Guccione, Julius M.

2013-01-01

Injection of Algisyl-LVR, a treatment under clinical development, is intended to treat patients with dilated cardiomyopathy. This treatment was recently used for the first time in patients who had symptomatic heart failure. In all patients, cardiac function of the left ventricle (LV) improved significantly, as manifested by consistent reduction of the LV volume and wall stress. Here we describe this novel treatment procedure and the methods used to quantify its effects on LV wall stress and function. Algisyl-LVR is a biopolymer gel consisting of Na+-Alginate and Ca2+-Alginate. The treatment procedure was carried out by mixing these two components and then combining them into one syringe for intramyocardial injections. This mixture was injected at 10 to 19 locations mid-way between the base and apex of the LV free wall in patients. Magnetic resonance imaging (MRI), together with mathematical modeling, was used to quantify the effects of this treatment in patients before treatment and at various time points during recovery. The epicardial and endocardial surfaces were first digitized from the MR images to reconstruct the LV geometry at end-systole and at end-diastole. Left ventricular cavity volumes were then measured from these reconstructed surfaces. Mathematical models of the LV were created from these MRI-reconstructed surfaces to calculate regional myofiber stress. Each LV model was constructed so that 1) it deforms according to a previously validated stress-strain relationship of the myocardium, and 2) the predicted LV cavity volume from these models matches the corresponding MRI-measured volume at end-diastole and end-systole. Diastolic filling was simulated by loading the LV endocardial surface with a prescribed end-diastolic pressure. Systolic contraction was simulated by concurrently loading the endocardial surface with a prescribed end-systolic pressure and adding active contraction in the myofiber direction. Regional myofiber stress at end-diastole and

Mercury mass measurement in fluorescent lamps via neutron activation analysis

Science.gov (United States)

Viererbl, L.; Vinš, M.; Lahodová, Z.; Fuksa, A.; Kučera, J.; Koleška, M.; Voljanskij, A.

2015-11-01

Mercury is an essential component of fluorescent lamps. Not all fluorescent lamps are recycled, resulting in contamination of the environment with toxic mercury, making measurement of the mercury mass used in fluorescent lamps important. Mercury mass measurement of lamps via instrumental neutron activation analysis (NAA) was tested under various conditions in the LVR-15 research reactor. Fluorescent lamps were irradiated in different positions in vertical irradiation channels and a horizontal channel in neutron fields with total fluence rates from 3×108 cm-2 s-1 to 1014 cm-2 s-1. The 202Hg(n,γ)203Hg nuclear reaction was used for mercury mass evaluation. Activities of 203Hg and others induced radionuclides were measured via gamma spectrometry with an HPGe detector at various times after irradiation. Standards containing an Hg2Cl2 compound were used to determine mercury mass. Problems arise from the presence of elements with a large effective cross section in luminescent material (europium, antimony and gadolinium) and glass (boron). The paper describes optimization of the NAA procedure in the LVR-15 research reactor with particular attention to influence of neutron self-absorption in fluorescent lamps.

Development of instrumentation and technological procedures for industrial and medical applications of ionizing radiation. Final project report, October 2001 - September 2004

International Nuclear Information System (INIS)

Osmera, B.; Posta, S.; Jansky, B.

2004-01-01

The project consisted of the following stages (Stage symbol; Stage name; Stage output; Fields of application): E01; Neutron detection and spectrometry using doped scintillators; A spectrometry system for fast neutron measurements in low neutron intensity fields; Fuel storage facilities, analysis of samples contaminated by uranium and transuranium elements, university teaching. E02; Neutron and gamma spectrometry in mixed neutron/gamma fields; A system for the spectrometry of mixed neutron/gamma fields; Neutron and gamma spectrometry in working environment, radiation protection of military equipment, metrology in neutron/gamma fields, university teaching. E03; Neutron detection methods based on fission fragment track counting; A system for neutron fluence determination based on fission fragment track counting; Storage cask monitoring, radiation protection, university teaching. E04; Calibration methods for gamma scanning of fuel assemblies; A calibration system for fuel pin inspection via gamma distribution measurement; Nuclear fuel burnup assessment, BUC technology implementation, spent fuel processing. E05; Optimization of the neutron capture therapy method at the LVR-15 reactor; Treatment of selected tumorous and non-tumorous diseases; Radiotherapy - novel therapy methods in oncology, novel methods of non-tumorous disease therapy, operation of the irradiation facilities of the LVR-15 reactor with simultaneous neutron capture therapy procedures. Details of each of the stages are given. (P.A.)

Mercury mass measurement in fluorescent lamps via neutron activation analysis

International Nuclear Information System (INIS)

Viererbl, L.; Vinš, M.; Lahodová, Z.; Fuksa, A.; Kučera, J.; Koleška, M.; Voljanskij, A.

2015-01-01

Mercury is an essential component of fluorescent lamps. Not all fluorescent lamps are recycled, resulting in contamination of the environment with toxic mercury, making measurement of the mercury mass used in fluorescent lamps important. Mercury mass measurement of lamps via instrumental neutron activation analysis (NAA) was tested under various conditions in the LVR-15 research reactor. Fluorescent lamps were irradiated in different positions in vertical irradiation channels and a horizontal channel in neutron fields with total fluence rates from 3×10 8 cm −2 s −1 to 10 14 cm −2 s −1 . The 202 Hg(n,γ) 203 Hg nuclear reaction was used for mercury mass evaluation. Activities of 203 Hg and others induced radionuclides were measured via gamma spectrometry with an HPGe detector at various times after irradiation. Standards containing an Hg 2 Cl 2 compound were used to determine mercury mass. Problems arise from the presence of elements with a large effective cross section in luminescent material (europium, antimony and gadolinium) and glass (boron). The paper describes optimization of the NAA procedure in the LVR-15 research reactor with particular attention to influence of neutron self-absorption in fluorescent lamps. - Highlights: • Mercury is an essential component of fluorescent lamps. • Fluorescent lamps were irradiated in neutron fields in research reactor. • 203 Hg induced radionuclide activity was measured using gamma spectrometry. • Mercury mass in fluorescent lamps can be measured by neutron activation analysis.

Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

International Nuclear Information System (INIS)

Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

2011-01-01

Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

Radiation parameter monitoring of the irradiation channel of the RVS-3 loop during the FRAMATOME 1 experiment in 1996/1997

International Nuclear Information System (INIS)

Hrabanek, J.

1997-11-01

The monitoring system serving to measure the neutron fluence rate with self-powered rhodium detectors and the radiative heating on iron with calorimeters is highlighted. The sensor signal transmission routes and instrumentation for their measurement and recording are described. The method of observed data evaluation is characterized and the results of this processing are given for the FRAMATOME 1 experiment, which was carried out on the RVS-3 loop of the LVR-15 reactor in 1996-1997. (author)

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

International Nuclear Information System (INIS)

Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit

2009-12-01

This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

Energy Technology Data Exchange (ETDEWEB)

Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit (Chalmers Univ. of Technology, Dept. of Nuclear Engineering, SE-412 96 Goeteborg (Sweden))

2009-12-15

This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

Determination of radioisotopic contents of high-active materials in non-standard conditions

International Nuclear Information System (INIS)

Klupak, V.; Viererbl, L.; Lahodova, Z.; Vins, M.

2014-01-01

The light-water research reactor LVR-15, operated by Research |Centre Rez, serves as a source of neutrons for applied and basic research. One of the main areas of applied research is the influence of neutron radiation field on the physical and mechanical properties of construction materials used in reactor technology. The presented contribution deals with the determination of a radioisotopic composition and the estimate of its activity.The determination is based on gamma spectrometry measurements. The purpose of the measurements was the determination of the main sources of radiation and the result was used to prepare a procedure for the handling, transport and testing of samples in the rigs. (authors)

ECP measurements under neutron and gamma ray in in-pile loop and their data evaluation by water radiolysis calculations

Energy Technology Data Exchange (ETDEWEB)

Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic)

2010-07-01

In order to establish reliable electrochemical corrosion potential (ECP) sensors for applying in reactor core peripherals of power plants, performance tests of sensors under irradiation were carried out in the in-pile loop of the experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Responses of different kinds of sensors under neutron and gamma irradiation conditions have been compared each other. Corrosive conditions along the in-pile loop were calculated by water radiolysis calculation code, WRAC-J and calculated corrosive conditions were compared with the measured results. As a result of the evaluation, it was confirmed that the ECP sensors could be applied to irradiation conditions of reactor peripherals, while the water radiolysis model could be also applied for evaluation of corrosive conditions of reactor peripherals. (author)

Determination of Trace Concentration in TMD Detectors using PGAA

Directory of Open Access Journals (Sweden)

Tomandl I.

2015-01-01

Full Text Available Transmutation detectors could be alternative to the traditional activation detector method for neutron fluence dosimetry at power nuclear reactors. This new method require an isotopically highly-sensitive, non-destructive in sense of compactness as well as isotopic content, precise and standardly used analytical method for trace concentration determination. The capability of Prompt Gamma-ray Activation Analysis (PGAA for determination of trace concentrations of transmuted stable nuclides in the metallic foils of Ni, Au, Cu and Nb, which were irradiated for 21 days in the reactor core at the LVR-15 research reactor in Řež, is reported. The PGAA measurements of these activation foils were performed at the PGAA facility at Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII in Garching.

Fischer-Tropsch synthesis in slurry-phase reactors using Co/SBA-15 catalysts

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, J.J.; Lima, L.A.; Lima, W.S.; Rodrigues, M.G.F. [Universidade Federal de Campina Grande (UFCG), PB (Brazil). Unidade Academica de Engenharia Quimica], e-mail: meiry@deq.ufcg.edu.br; Fernandes, F.A.N. [Universidade Federal do Ceara (UFCE), CE (Brazil). Dept. de Engenharia Quimica

2011-07-15

The objective of this work is to describe the production of bifunctional catalysts using the incipient humidity method, producing catalysts with 15 wt.% cobalt supported in SBA-15 molecular sieve, to be applied in the Fischer-Tropsch (FT) reaction. The originality of this work is its focus on the use of a 15 wt.% Co/SBA-15 catalyst in FT synthesis in slurry reactors. The deposition of cobalt over SBA-15 support was accomplished by impregnation with a 0.1-M aqueous solution of cobalt nitrate. The Fischer-Tropsch synthesis was carried out with the catalyst at 240 deg C and 20 atm, under a COH{sub 2} atmosphere (molar ratio= 1), in a slurry reactor for 8 hours. X-ray diffraction measurements showed that the calcined cobalt catalyst did not modify the structure of SBA-15, proving that Co was present under the form of Co{sub 3}O{sub 4} in the catalyst. The addition of cobalt in the SBA-15 decreased the specific superficial area of the molecular sieve. The 15 wt.% Co/SBA-15 catalyst had a 40% CO conversion rate and a high selectivity towards the production of C{sub 5}{sup +} (53.9% after 8 hours). (author)

HECTR [Hydrogen Event Containment Transient Response] Version 1.5N: A modification of HECTR Version 1.5 for application to N Reactor

International Nuclear Information System (INIS)

Camp, A.L.; Dingman, S.E.

1987-05-01

This report describes HECTR Version 1.5N, which is a special version of HECTR developed specifically for application to the N Reactor. HECTR is a fast-running, lumped-parameter containment analysis computer program that is most useful for performing parametric studies. The main purpose of HECTR is to analyze nuclear reactor accidents involving the transport and combustion of hydrogen, but HECTR can also function as an experiment analysis tool and can solve a limited set of other types of containment problems. Version 1.5N is a modification of Version 1.5 and includes changes to the spray actuation logic, and models for steam vents, vacuum breakers, and building cross-vents. Thus, all of the key features of the N Reactor confinement can be modeled. HECTR is designed for flexibility and provides for user control of many important parameters, if built-in correlations and default values are not desired

The AMPS 1.5 MW low-pressure compact reactor

International Nuclear Information System (INIS)

Hewitt, J.S.

1987-01-01

The 1.5-MWt reactor of the Autonomous Marine Power Source (AMPS) is designed to meet the unusual requirements of its first application. To provide for 100 kWe (net) on board self-sustaining manned submersible vehicles, the AMPS reactor must deliver safely, reliably and without direct operator surveillance, its thermal output to freon Rankine-cycle engines at thermodynamically useful temperatures. It must also conform to space and weight limits on the order of less than 50 cubic metres and 70 tonnes. The safety requirements are met by (i) limiting lifetime excess reactivity requirements by incorporation of burnable poison in the U-Zr-H fuel, (ii) maintaining nominal pressures in the light-water primary system at about 1 atmosphere, and (iii) maintaining a large volume of primary reserve coolant at temperature depressed relative to that of the circulating coolant. The latter averages 90 degrees celsius as it is pumped around loops that include the reactor core and the freon evaporators during normal operation. In the event of loss of pumped flow, the system defaults by intrinsic means to core cooling through natural convective exchange with the reserve coolant. In the post-shutdown situation, this passive cooling mode continues to operate regardless of vessel orientation and decay heat is safely dissipated to the sea. The design of the AMPS system, including the reactor, the freon engines, the control and monitoring system, the safety shut-down system and the power source container, are in advanced stages of design. (author)

The Influence of RSG-GAS Primary Pump Operation Concerning the Rise Water Level of Reactor Pool in 15 MW Reactor Power

International Nuclear Information System (INIS)

Djunaidi

2004-01-01

The expansion of air volume in the delay chamber shows in rise water level of reactor pool during the operation. The rises of water level in the reactor pool is not quite from the expansion of air volume in the delay chamber, but some influence the primary pump operation. The purpose evaluated of influence primary pump is to know the influence primary pump power concerning the rise water level during the reactor operation. From the data collection during 15 MW power operation in the last core 42 the influence of primary pump operation concerning the rise water level in the reactor pool is 34.48 % from the total increased after operation during 12 days. (author)

Radiation stability tests of the SMELS materials in the LVR-15 experimental reactor at Řež

Czech Academy of Sciences Publication Activity Database

Kučera, Jan; Eguskiza, M.; De Corte, F.; Rabouch, P.

2003-01-01

Roč. 257, č. 3 (2003), s. 677-682 ISSN 0236-5731 R&D Projects: GA ČR GA202/03/0891 Institutional research plan: CEZ:AV0Z1048901 Keywords : standardization * neutron activation analysis Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.472, year: 2003

Effect of Drawer Master Modeling of ZPPR15 Phase A Reactor Physics Experiment on Integral Parameter

International Nuclear Information System (INIS)

Yoo, Jae Woon; Kim, Sang Ji

2011-01-01

As a part of an International-Nuclear Engineering Research Initiative (I-NERI) Project, KAERI and ANL are analyzing the ZPPR-15 reactor physics experiments. The ZPPR-15 experiments were carried out in support of the Integral Fast Reactor (IFR) project. Because of lack of the experimental data, verifying and validating the core neutronics analysis code for metal fueled sodium cooled fast reactors (SFR) has been one of the big concerns. KAERI is developing the metal fuel loaded SFR and plans to construct the demonstration SFR by around 2028. Database built through this project and its result of analysis will play an important role in validating the SFR neutronics characteristics. As the first year work of I-NERI project, KAERI analyzed ZPPR-15 Phase A experiment among four phases (Phase A to D). The effect of a drawer master modeling on the integral parameter was investigated. The approximated benchmark configurations for each loading were constructed to be used for validating a deterministic code

Upgrade for the epithermal neutron beam at NRI Rez

International Nuclear Information System (INIS)

Marek, M.; Flibor, S.; Viererbl, L.; Burian, J.; Rejchrt, J.; Klupak, V.; Gambarini, G.; Vanossi, E.

2006-01-01

The epithermal neutron beam facility designed for pre-clinical neutron capture therapy research has been operated at LVR-15 reactor for more than ten years. The construction of the beam filter has been recently modified especially for the shielding quality of the beam shutter to be improved. The parameters of the upgraded beam were calculated with the MCNP code and a new source term for the NCTPLAN treatment planning software was evaluated. The calculated source term was consequently scaled according to the results of measurements in the free beam and in the 50x50x25 cm 3 water phantom. (author)

New developments in instrumentation for strain scanning in NPI Rez

CERN Document Server

Vrána, M; Lukás, P; Dubsky, J; Wagner, V

2000-01-01

Two high-resolution neutron diffractometers dedicated to strain /stress investigations in polycrystalline materials are routinely operated at the medium-power reactor LVR-15 in Rez and are also opened for external users. Good luminosity of the diffractometers and a sufficiently high-resolution (FWHM of the instrumental Delta d/d - profile can be less than 2*10/sup -3/ at d/sub hkl/=0.2 nm) permit investigations of both the macro- and microstrains in the sample gauge volumes of several cubic millimetres. Excellent properties and abilities of the neutron diffraction strain scanners are documented by some results of experimental measurements. (11 refs).

Phenotypic and molecular differences between rats selectively bred to voluntarily run high vs. low nightly distances.

Science.gov (United States)

Roberts, Michael D; Brown, Jacob D; Company, Joseph M; Oberle, Lauren P; Heese, Alexander J; Toedebusch, Ryan G; Wells, Kevin D; Cruthirds, Clayton L; Knouse, John A; Ferreira, J Andries; Childs, Thomas E; Brown, Marybeth; Booth, Frank W

2013-06-01

The purpose of the present study was to partially phenotype male and female rats from generations 8-10 (G8-G10) that had been selectively bred to possess low (LVR) vs. high voluntary running (HVR) behavior. Over the first 6 days with wheels, 34-day-old G8 male and female LVRs ran shorter distances (P HVR counterparts, respectively. HVR and LVR lines consumed similar amounts of standard chow with or without wheels. No inherent difference existed in PGC-1α mRNA in the plantaris and soleus muscles of LVR and HVR nonrunners, although G8 LVR rats inherently possessed less NADH-positive superficial plantaris fibers compared with G8 HVR rats. While day 28 body mass tended to be greater in both sexes of G9-G10 LVR nonrunners vs. G9-G10 HVR nonrunners (P = 0.06), body fat percentage was similar between lines. G9-G10 HVRs had fat mass loss after 6 days of running compared with their prerunning values, while LVR did not lose or gain fat mass during the 6-day voluntary running period. RNA deep sequencing efforts in the nucleus accumbens showed only eight transcripts to be >1.5-fold differentially expressed between lines in HVR and LVR nonrunners. Interestingly, HVRs presented less Oprd1 mRNA, which ties in to potential differences in dopaminergic signaling between lines. This unique animal model provides further evidence as to how exercise may be mechanistically regulated.

15 N utilization in nitride nuclear fuels for advanced nuclear power reactors and accelerator - driven systems

International Nuclear Information System (INIS)

Axente, D.

2005-01-01

15 N utilization for nitride nuclear fuels production for nuclear power reactors and accelerator - driven systems is presented. Nitride nuclear fuel is the obvious choice for advanced nuclear reactors and ADS because of its favorable properties: a high melting point, excellent thermal conductivity, high fissile density, lower fission gas release and good radiation tolerance. The application of nitride fuels in nuclear reactors and ADS requires use of 15 N enriched nitrogen to suppress 14 C production due to (n,p) reaction on 14 N. Accelerator - driven system is a recent development merging of accelerator and fission reactor technologies to generate electricity and transmute long - lived radioactive wastes as minor actinides: Np, Am, Cm. A high-energy proton beam hitting a heavy metal target produces neutrons by spallation. The neutrons cause fission in the fuel, but unlike in conventional reactors, the fuel is sub-critical and fission ceases when the accelerator is turned off. Nitride fuel is a promising candidate for transmutation in ADS of minor actinides, which are converted into nitrides with 15 N for that purpose. Tacking into account that the world wide market is about 20 to 40 Kg 15 N annually, the supply of that isotope for nitride fuel production for nuclear power reactors and ADS would therefore demand an increase in production capacity by a factor of 1000. For an industrial plant producing 100 t/y 15 N, using present technology of isotopic exchange in NITROX system, the first separation stage of the cascade would be fed with 10M HNO 3 solution of 600 mc/h flow - rate. If conversion of HNO 3 into NO, NO 2 , at the enriching end of the columns, would be done with gaseous SO 2 , for a production plant of 100 t/y 15 N a consumption of 4 million t SO 2 /y and a production of 70 % H 2 SO 4 waste solution of 4.5 million mc/y are estimated. The reconversion of H 2 SO 4 into SO 2 in order to recycle of SO 2 is a problem to be solved to compensate the cost of SO 2

Phenotypic and molecular differences between rats selectively bred to voluntarily run high vs. low nightly distances

Science.gov (United States)

Roberts, Michael D.; Brown, Jacob D.; Company, Joseph M.; Oberle, Lauren P.; Heese, Alexander J.; Toedebusch, Ryan G.; Wells, Kevin D.; Cruthirds, Clayton L.; Knouse, John A.; Ferreira, J. Andries; Childs, Thomas E.; Brown, Marybeth

2013-01-01

The purpose of the present study was to partially phenotype male and female rats from generations 8–10 (G8–G10) that had been selectively bred to possess low (LVR) vs. high voluntary running (HVR) behavior. Over the first 6 days with wheels, 34-day-old G8 male and female LVRs ran shorter distances (P HVR counterparts, respectively. HVR and LVR lines consumed similar amounts of standard chow with or without wheels. No inherent difference existed in PGC-1α mRNA in the plantaris and soleus muscles of LVR and HVR nonrunners, although G8 LVR rats inherently possessed less NADH-positive superficial plantaris fibers compared with G8 HVR rats. While day 28 body mass tended to be greater in both sexes of G9–G10 LVR nonrunners vs. G9–G10 HVR nonrunners (P = 0.06), body fat percentage was similar between lines. G9–G10 HVRs had fat mass loss after 6 days of running compared with their prerunning values, while LVR did not lose or gain fat mass during the 6-day voluntary running period. RNA deep sequencing efforts in the nucleus accumbens showed only eight transcripts to be >1.5-fold differentially expressed between lines in HVR and LVR nonrunners. Interestingly, HVRs presented less Oprd1 mRNA, which ties in to potential differences in dopaminergic signaling between lines. This unique animal model provides further evidence as to how exercise may be mechanistically regulated. PMID:23552494

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Electrochemical Oxidation of Propene with a LSF15/CGO10 Electrochemical Reactor

DEFF Research Database (Denmark)

Ippolito, Davide; Kammer Hansen, Kent

2014-01-01

A porous electrochemical reactor, made of La0.85Sr0.15FeO3 (LSF) as electrode and Ce0.9Gd0.1O1.95 (CGO) as electrolyte, was studied for the electrochemical oxidation of propene over a wide range of temperatures. Polarization was found to enhance propene oxidation rate. Ce0.9Gd0.1O1.95 was used...... as infiltration material to enhance the effect of polarization on propene oxidation rate, especially at low temperatures. The influence of infiltrated material, as a function of heat treatment, on the reactor electrochemical behavior has been evaluated by using electrochemical impedance spectroscopy...... in suppressing the competing oxygen evolution reaction and promoting the oxidation of propene under polarization, with faradaic efficiencies above 70% at 250◦C. © 2014 The Electrochemical Society....

Volatility of V15Cr5Ti fusion reactor alloy

International Nuclear Information System (INIS)

Neilson, R.M. Jr.

1986-01-01

One potential hazard from the presence of activation products in fusion facilities is accidental oxidation-driven volatility of those activation products. Scoping experiments were conducted to investigate the oxidation and elemental volatility of candidate fusion reactor alloy V15Cr5Ti as a function of time, temperature, and test atmosphere. Experiments in air and in argon carrier gases containing 10 4 to 10 1 Pa (10 -1 to 10 -4 atm) oxygen were conducted to investigate the lower oxygen partial pressure limit for the formation of a low melting point (approximately 650 0 C), high volatility, oxide layer and its formation rate. Experiments to determine the elemental volatility of alloy constituents in air at temperatures of 700 0 C to greater than 1600 0 C. Some of these volatility experiments used V15Cr5Ti that was arc-remelted to incorporate small quantities (<0.1 wt. %) of Sc and Ca. Incorporation of Sc and Ca in test specimens permitted volatility measurement of radioactive constituents present only after activation of V15Cr5Ti

Development of the MCNPX model for the portable HPGe detector

International Nuclear Information System (INIS)

Koleska, Michal; Viererbl, Ladislav; Marek, Milan

2014-01-01

The portable HPGe coaxial detector Canberra Big MAC is used in LVR-15 research reactor for spectrometric measurement of spent nuclear fuel. The fuel is measured in the dedicated system located in the spent fuel pool situated near the reactor. For the purpose of the spectrometric system calibration, the detector was precisely modeled with the MCNPX code. This model was constructed with the data acquired from the technical specification provided by the manufacturer and from the data obtained by the radiography of the crystal. The detector model was verified on the experimental data measured with available standard radionuclide sources and on-site prepared 110m Ag source. - Highlights: • Inner structure of the HPGe detector is determined. • An MCNPX model of the detector is developed. • The model is verified using different sources for two measurement geometries

Temper-bead repair-welding of neutron-irradiated reactor (pressure) vessel by low-heat-input TIG and YAG laser welding

International Nuclear Information System (INIS)

Nakata, Kiyotomo; Ozawa, Masayoshi; Kamo, Kazuhiko

2006-01-01

Weldability in neutron-irradiated low alloy steel for reactor (pressure) vessel has been studied by temper-bead repair-welding of low-heat-input TIG and YAG laser welding. A low alloy steel and its weld, and stainless steel clad and nickel (Ni)-based alloy clad were irradiated in a materials test reactor (LVR-15, Czech Republic) up to 1.4 x 10 24 n/m 2 (>1 MeV) at 290degC, which approximately corresponds to the maximum neutron fluence of 60-year-operation plants' vessels. The He concentration in the irradiated specimens was estimated to be up to 12.9 appm. The repair-welding was carried out by TIG and YAG laser welding at a heat input from 0.06 to 0.86 MJ/m. The mechanical tests of tensile, impact, side bend and hardness were carried out after the repair-welding. Cracks were not observed in the irradiated low alloy steel and its weld by temper-bead repair-welding. Small porosities were formed in the first and second layers of the repair-welds of low alloy steel (base metal). However, only a few porosities were found in the repair-welds of the weld of low alloy steel. From the results of mechanical tests, the repair-welding could be done in the irradiated weld of low alloy steel containing a He concentration up to 12.9 appm, although repair-welding could be done in base metal of low alloy steel containing up to only 1.7 appmHe. On the other hand, cracks occurred in the heat affected zones of stainless steel and Ni-based alloy clads by repair-welding, except by YAG laser repair-welding at a heat input of 0.06 MJ/m in stainless steel clad containing 1.7 appmHe. Based on these results, the determination processes were proposed for optimum parameters of repair-welding of low alloy steel and clad used for reactor (pressure) vessel. (author)

Research reactor utilization. Summary reports of three study group meetings: Irradiation techniques at research reactors, held in Istanbul 15-19 November 1965; Research reactor operation and maintenance problems, held in Caracas 6-10 December 1965; and Research reactor utilization in the Far East, held in Lucas Heights 28 February - 4 March 1966

International Nuclear Information System (INIS)

1967-01-01

The three sections of this book, which are summary reports of three Study Group meetings of the IAEA: Irradiation techniques at research reactors, Istanbul, 15-19 November 1965; Research reactor operation and maintenance problems, Caracas, 6-10 December 1965; and Research reactor utilization in the Far East, Lucas Heights, Australia, 28 February - 4 March 1966. These meetings were the latest in a series designed to promote efficient utilization of research reactors, to disseminate information on advances in techniques, to discuss common problems in reactor operations, and to outline some advanced areas of reactor-based research. (author)

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Fission product chemistry in severe nuclear reactor accidents, specialists' meeting at JRC-Ispra, 15-17 January 1990

International Nuclear Information System (INIS)

Nichols, A.L.

1990-05-01

A specialists' meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions). (author)

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

3D Echo pilot study of geometric left ventricular changes after acute myocardial infarction.

Science.gov (United States)

Vieira, Marcelo Luiz Campos; Oliveira, Wercules Antonio; Cordovil, Adriana; Rodrigues, Ana Clara Tude; Mônaco, Cláudia Gianini; Afonso, Tânia; Lira Filho, Edgar Bezerra; Perin, Marco; Fischer, Cláudio Henrique; Morhy, Samira Saady

2013-07-01

Left ventricular remodeling (LVR) after AMI characterizes a factor of poor prognosis. There is little information in the literature on the LVR analyzed with three-dimensional echocardiography (3D ECHO). To analyze, with 3D ECHO, the geometric and volumetric modifications of the left ventricle (VE) six months after AMI in patients subjected to percutaneous primary treatment. Prospective study with 3D ECHO of 21 subjects (16 men, 56 ± 12 years-old), affected by AMI with ST segment elevation. The morphological and functional analysis (LV) with 3D ECHO (volumes, LVEF, 3D sphericity index) was carried out up to seven days and six months after the AMI. The LVR was considered for increase > 15% of the end diastolic volume of the LV (LVEDV) six months after the AMI, compared to the LVEDV up to seven days from the event. Eight (38%) patients have presented LVR. Echocardiographic measurements (n = 21 patients): I- up to seven days after the AMI: 1- LVEDV: 92.3 ± 22.3 mL; 2- LVEF: 0.51 ± 0.01; 3- sphericity index: 0.38 ± 0.05; II- after six months: 1- LVEDV: 107.3 ± 26.8 mL; 2- LVEF: 0.59 ± 0.01; 3- sphericity index: 0.31 ± 0.05. Correlation coefficient (r) between the sphericity index up to seven days after the AMI and the LVEDV at six months (n = 8) after the AMI: r: 0.74, p = 0.0007; (r) between the sphericity index six months after the AMI and the LVEDV at six months after the AMI: r: 0.85, p series, LVR has been observed in 38% of the patients six months after the AMI. The three-dimensional sphericity index has been associated to the occurrence of LVR.

Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

International Nuclear Information System (INIS)

Lai, W.; Collins, E.K.

1977-01-01

This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 15, 2011, 10:00 PM status

International Nuclear Information System (INIS)

2011-01-01

This situation note is established according to the information gained on March 15, 2011, at 10:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 and of the spent fuel pools of reactors No. 4, 5 and 6 of the Fukushima I site (Dai-ichi), 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.)

Cocaine self-administration and reinstatement in female rats selectively bred for high and low voluntary running.

Science.gov (United States)

Smethells, J R; Zlebnik, N E; Miller, D K; Will, M J; Booth, F; Carroll, M E

2016-10-01

Previous research has found that rats behaviorally screened for high (vs. low) wheel running were more vulnerable to cocaine abuse. To assess the extent to which a genetic component is involved in this drug-abuse vulnerability, rats selectively bred for high or low voluntary running (HVR or LVR, respectively) were examined for differences in cocaine seeking in the present study. Female rats were trained to lever press for food and then were assessed for differences in acquisition of cocaine (0.4mg/kg; i.v.) self-administration across 10 sessions. Once acquired, rats self-administered cocaine for a 14-day maintenance phase, followed by a 14-day extinction phase when cocaine was no longer available. Subsequently, reinstatement of cocaine seeking was examined with priming injections of cocaine (5, 10 & 15mg/kg), caffeine (30mg/kg), yohimbine (2.5mg/kg) and cocaine-paired cues. A greater percentage of LVR rats met the acquisition criteria for cocaine self-administration and in fewer sessions than HVR rats. No differences in responding for cocaine were observed between phenotypes during maintenance. However, during extinction LVR rats initially responded at higher rates and persisted in cocaine seeking for a greater number of sessions. No phenotype differences were observed following drug and cue-primed reinstatement of cocaine seeking. In general, LVR rats were more sensitive to the reinforcing effects of cocaine than HVR rats during periods of transition into and out of cocaine self-administration. Thus, LVR rats sometimes showed a greater vulnerability cocaine seeking than HVR rats. Published by Elsevier Ireland Ltd.

Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

International Nuclear Information System (INIS)

Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

2011-01-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

Lung volume reduction of pulmonary emphysema: the radiologist task.

Science.gov (United States)

Milanese, Gianluca; Silva, Mario; Sverzellati, Nicola

2016-03-01

Several lung volume reduction (LVR) techniques have been increasingly evaluated in patients with advanced pulmonary emphysema, especially in the last decade. Radiologist plays a pivotal role in the characterization of parenchymal damage and, thus, assessment of eligibility criteria. This review aims to discuss the most common LVR techniques, namely LVR surgery, endobronchial valves, and coils LVR, with emphasis on the role of computed tomography (CT). Several trials have recently highlighted the importance of regional quantification of emphysema by computerized CT-based segmentation of hyperlucent parenchyma, which is strongly recommended for candidates to any LVR treatment. In particular, emphysema distribution pattern and fissures integrity are evaluated to tailor the choice of the most appropriate LVR technique. Furthermore, a number of CT measures have been tested for the personalization of treatment, according to imaging detected heterogeneity of parenchymal disease. CT characterization of heterogeneous parenchymal abnormalities provides criteria for selection of the preferable treatment in each patient and improves outcome of LVR as reflected by better quality of life, higher exercise tolerance, and lower mortality.

ECP measurements in the BWR-1 water loop relative to water composition changes

Energy Technology Data Exchange (ETDEWEB)

Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

2010-07-01

The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

Low Vision Depression Prevention Trial in Age-Related Macular Degeneration

Science.gov (United States)

Rovner, Barry W.; Casten, Robin J.; Hegel, Mark T.; Massof, Robert W.; Leiby, Benjamin E.; Ho, Allen C.; Tasman, William S.

2014-01-01

Purpose To compare the efficacy of behavior activation (BA) + low vision rehabilitation (LVR) with supportive therapy (ST) + LVR to prevent depressive disorders in patients with age-related macular degeneration (AMD). Design Single-masked, attention-controlled, randomized, clinical trial with outcome assessment at 4 months. Participants Patients with AMD and subsyndromal depressive symptoms attending retina practices (n = 188). Interventions Before randomization, all subjects had 2 outpatient LVR visits, and were then randomized to in-home BA+LVR or ST+LVR. Behavior activation is a structured behavioral treatment that aims to increase adaptive behaviors and achieve valued goals. Supportive therapy is a nondirective, psychological treatment that provides emotional support and controls for attention. Main Outcome Measures The Diagnostic and Statistical Manual IV defined depressive disorder based on the Patient Health Questionnaire-9 (primary outcome), Activities Inventory, National Eye Institute Vision Function Questionnaire–25 plus Supplement (NEI-VFQ), and NEI-VFQ quality of life (secondary outcomes). Results At 4 months, 11 BA+LVR subjects (12.6%) and 18 ST+LVR subjects (23.4%) developed a depressive disorder (relative risk [RR], 0.54; 95% CI, 0.27–1.06; P = 0.067). In planned adjusted analyses the RR was 0.51 (95% CI, 0.27–0.98; P = 0.04). A mediational analysis suggested that BA+LVR prevented depression to the extent that it enabled subjects to remain socially engaged. In addition, BA+LVR was associated with greater improvements in functional vision than ST+LVR, although there was no significant between-group difference. There was no significant change or between-group difference in quality of life. Conclusions An integrated mental health and low vision intervention halved the incidence of depressive disorders relative to standard outpatient LVR in patients with AMD. As the population ages, the number of persons with AMD and the adverse effects of comorbid

Low vision depression prevention trial in age-related macular degeneration: a randomized clinical trial.

Science.gov (United States)

Rovner, Barry W; Casten, Robin J; Hegel, Mark T; Massof, Robert W; Leiby, Benjamin E; Ho, Allen C; Tasman, William S

2014-11-01

To compare the efficacy of behavior activation (BA) + low vision rehabilitation (LVR) with supportive therapy (ST) + LVR to prevent depressive disorders in patients with age-related macular degeneration (AMD). Single-masked, attention-controlled, randomized, clinical trial with outcome assessment at 4 months. Patients with AMD and subsyndromal depressive symptoms attending retina practices (n = 188). Before randomization, all subjects had 2 outpatient LVR visits, and were then randomized to in-home BA+LVR or ST+LVR. Behavior activation is a structured behavioral treatment that aims to increase adaptive behaviors and achieve valued goals. Supportive therapy is a nondirective, psychological treatment that provides emotional support and controls for attention. The Diagnostic and Statistical Manual IV defined depressive disorder based on the Patient Health Questionnaire-9 (primary outcome), Activities Inventory, National Eye Institute Vision Function Questionnaire-25 plus Supplement (NEI-VFQ), and NEI-VFQ quality of life (secondary outcomes). At 4 months, 11 BA+LVR subjects (12.6%) and 18 ST+LVR subjects (23.4%) developed a depressive disorder (relative risk [RR], 0.54; 95% CI, 0.27-1.06; P = 0.067). In planned adjusted analyses the RR was 0.51 (95% CI, 0.27-0.98; P = 0.04). A mediational analysis suggested that BA+LVR prevented depression to the extent that it enabled subjects to remain socially engaged. In addition, BA+LVR was associated with greater improvements in functional vision than ST+LVR, although there was no significant between-group difference. There was no significant change or between-group difference in quality of life. An integrated mental health and low vision intervention halved the incidence of depressive disorders relative to standard outpatient LVR in patients with AMD. As the population ages, the number of persons with AMD and the adverse effects of comorbid depression will increase. Promoting interactions between ophthalmology, optometry

The lung volume reduction coil for the treatment of emphysema : a new therapy in development

NARCIS (Netherlands)

Klooster, Karin; ten Hacken, Nick H. T.; Slebos, Dirk-Jan

Lung volume reduction (LVR) coil treatment is a novel therapy for patients with severe emphysema. In this bilateral bronchoscopic treatment, approximately 10 LVR coils per lobe are delivered under fluoroscopic guidance in two sequential procedures. The LVR coil reduces lung volume by compressing the

Reactor outage schedule (tentative)

Energy Technology Data Exchange (ETDEWEB)

Walton, R.P.

1969-10-15

This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

Reactor outage schedule (tentative)

Energy Technology Data Exchange (ETDEWEB)

Walton, R.P.

1969-09-15

This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

Large-Scale Variations in Lumber Value Recovery of Yellow Birch and Sugar Maple in Quebec, Canada.

Science.gov (United States)

Hassegawa, Mariana; Havreljuk, Filip; Ouimet, Rock; Auty, David; Pothier, David; Achim, Alexis

2015-01-01

Silvicultural restoration measures have been implemented in the northern hardwoods forests of southern Quebec, Canada, but their financial applicability is often hampered by the depleted state of the resource. To help identify sites most suited for the production of high quality timber, where the potential return on silvicultural investments should be the highest, this study assessed the impact of stand and site characteristics on timber quality in sugar maple (Acer saccharum Marsh.) and yellow birch (Betula alleghaniensis Britt.). For this purpose, lumber value recovery (LVR), an estimate of the summed value of boards contained in a unit volume of round wood, was used as an indicator of timber quality. Predictions of LVR were made for yellow birch and sugar maple trees contained in a network of more than 22000 temporary sample plots across the Province. Next, stand-level variables were selected and models to predict LVR were built using the boosted regression trees method. Finally, the occurrence of spatial clusters was verified by a hotspot analysis. Results showed that in both species LVR was positively correlated with the stand age and structural diversity index, and negatively correlated with the number of merchantable stems. Yellow birch had higher LVR in areas with shallower soils, whereas sugar maple had higher LVR in regions with deeper soils. The hotspot analysis indicated that clusters of high and low LVR exist across the province for both species. Although it remains uncertain to what extent the variability of LVR may result from variations in past management practices or in inherent site quality, we argue that efforts to produce high quality timber should be prioritized in sites where LVR is predicted to be the highest.

Operational performance of the development of a 15 kW parabolic trough mid-temperature solar receiver/reactor for hydrogen production

International Nuclear Information System (INIS)

Hong, Hui; Liu, Qibin; Jin, Hongguang

2012-01-01

Highlights: ► A 15 kW solar chemical receiver/reactor for hydrogen production was developed. ► The solar thermochemical efficiency of the receiver/reactor was in the range of 20–28%. ► Hydrogen production exceeding 80% was achieved. ► The research results extend the application of mid-temperature solar thermal energy. -- Abstract: In this paper, we report the operational performance and energy conversion efficiency of a developed 15 kW solar chemical receiver/reactor for hydrogen production. A concentrated solar heat of around 200–300 °C was utilized to provide process heat to drive methanol steam reforming. A modified 15 kW direct-irradiation solar reactor coupled with a linear receiver positioned along the focal line of a one-axis parabolic trough concentrator was used. The experiments were conducted from 200 to 300 °C under a mean solar flux of 300–800 W/m 2 and a reactant feeding rate of 6 kg/h. Reactants were continuously fed, and the attained conversion rate of methanol was more than 70% at 700 W/m 2 . The typical solar thermochemical efficiency of solar thermal energy converted into chemical energy was in the 20–28% range. The overall energy efficiency of input solar power conversion into chemical energy reached up to 17% and may be further increased by improving solar field efficiency. Hydrogen production exceeding 80% was achieved. In addition, preliminary economic evaluation was performed, and methods for further improvement were proposed. This paper proves that solar hydrogen production is feasible by combining solar thermal energy with alternative fuel at around 200–300 °C, which is much lower than the temperature of other solar thermochemical processes. This may offer an economic approach to solar fuel production and extend the application of mid-temperature solar thermal energy.

Loop capabilities in Rez for water chemistry and corrosion control of cladding and in-core components

International Nuclear Information System (INIS)

Kysela, J.; Zmitko, M.; Srank, J.; Vsolak, R.

1999-01-01

Main characteristics of LVR-15 research reactor and its irradiation facilities are presented. For testing of cladding, internals and RPV materials specialised loop are used. There are now five high pressure loops modelling PWR, WWER or BWR water environment and chemistry. Loops can be connected with instrumented in-pile channels enable slow strain rate testing, 1CT or 2CT specimens loading and electrically heated rods exposition. Reactor dosimetry including neutronic parameters measurements and calculations and mock-up experiments are used. Water chemistry control involves gas (O 2 , H 2 ) dosing system, Orbisphere H 2 /O 2 measurement, electrochemical potential (ECP) measurements and specialised analytical chemistry laboratory. For cladding corrosion studies in-pile channels with four electrically heated rods with heat flux up to 100 W/cm 2 , void fraction 5 % at the outlet, inlet temperature 320 deg. C and flow velocity 3 m/s were development and tested. For corrosion layer investigation there is eddy current measurements and PIE techniques which use crud thickness measurement, chemical analyses of the crud, optical metallography, hydrogen analysis, SEM and TEM. (author)

Reactors of the world

International Nuclear Information System (INIS)

1971-01-01

Basic data relating to 127 power reactors in 15 countries which are expected to be in operation at the end of this year, with a total installed electrical generating capacity of 35 340.15 MW(e), and a listing of 361 research reactors in 46 countries are given in the 1971 edition of the IAEA handbook, Power and Research Reactors in Member States, which has just been published. This edition, the fourth, was prepared especially for the Fourth International Conference on the Peaceful Uses of Atomic Energy. (author)

Large-Scale Variations in Lumber Value Recovery of Yellow Birch and Sugar Maple in Quebec, Canada.

Directory of Open Access Journals (Sweden)

Mariana Hassegawa

Full Text Available Silvicultural restoration measures have been implemented in the northern hardwoods forests of southern Quebec, Canada, but their financial applicability is often hampered by the depleted state of the resource. To help identify sites most suited for the production of high quality timber, where the potential return on silvicultural investments should be the highest, this study assessed the impact of stand and site characteristics on timber quality in sugar maple (Acer saccharum Marsh. and yellow birch (Betula alleghaniensis Britt.. For this purpose, lumber value recovery (LVR, an estimate of the summed value of boards contained in a unit volume of round wood, was used as an indicator of timber quality. Predictions of LVR were made for yellow birch and sugar maple trees contained in a network of more than 22000 temporary sample plots across the Province. Next, stand-level variables were selected and models to predict LVR were built using the boosted regression trees method. Finally, the occurrence of spatial clusters was verified by a hotspot analysis. Results showed that in both species LVR was positively correlated with the stand age and structural diversity index, and negatively correlated with the number of merchantable stems. Yellow birch had higher LVR in areas with shallower soils, whereas sugar maple had higher LVR in regions with deeper soils. The hotspot analysis indicated that clusters of high and low LVR exist across the province for both species. Although it remains uncertain to what extent the variability of LVR may result from variations in past management practices or in inherent site quality, we argue that efforts to produce high quality timber should be prioritized in sites where LVR is predicted to be the highest.

Comparison of Physician-Predicted to Measured Low Vision Outcomes

Science.gov (United States)

Chan, Tiffany L.; Goldstein, Judith E.; Massof, Robert W.

2013-01-01

Purpose To compare low vision rehabilitation (LVR) physicians’ predictions of the probability of success of LVR to patients’ self-reported outcomes after provision of usual outpatient LVR services; and to determine if patients’ traits influence physician ratings. Methods The Activity Inventory (AI), a self-report visual function questionnaire, was administered pre and post-LVR to 316 low vision patients served by 28 LVR centers that participated in a collaborative observational study. The physical component of the Short Form-36, Geriatric Depression Scale, and Telephone Interview for Cognitive Status were also administered pre-LVR to measure physical capability, depression and cognitive status. Following patient evaluation, 38 LVR physicians estimated the probability of outcome success (POS), using their own criteria. The POS ratings and change in functional ability were used to assess the effects of patients’ baseline traits on predicted outcomes. Results A regression analysis with a hierarchical random effects model showed no relationship between LVR physician POS estimates and AI-based outcomes. In another analysis, Kappa statistics were calculated to determine the probability of agreement between POS and AI-based outcomes for different outcome criteria. Across all comparisons, none of the kappa values were significantly different from 0, which indicates the rate of agreement is equivalent to chance. In an exploratory analysis, hierarchical mixed effects regression models show that POS ratings are associated with information about the patient’s cognitive functioning and the combination of visual acuity and functional ability, as opposed to visual acuity or functional ability alone. Conclusions Physicians’ predictions of LVR outcomes appear to be influenced by knowledge of patients’ cognitive functioning and the combination of visual acuity and functional ability - information physicians acquire from the patient’s history and examination. However

Low-volume resuscitation using polyethylene glycol-20k in a preclinical porcine model of hemorrhagic shock.

Science.gov (United States)

Plant, Valerie; Limkemann, Ashley; Liebrecht, Loren; Blocher, Charles; Ferrada, Paula; Aboutanos, Michel; Mangino, Martin J

2016-12-01

Polyethylene glycol-20k (PEG-20k) is highly effective for low-volume resuscitation (LVR) by increasing tolerance to the low-volume state. In our rodent shock model, PEG-20k increased survival and expanded the "golden hour" 16-fold compared to saline. The molecular mechanism is largely attributed to normalizations in cell and tissue fluid shifts after low-flow ischemia resulting in efficient microvascular exchange. The objective of this study was to evaluate PEG-20k as an LVR solution for hemorrhagic shock in a preclinical model. Anesthetized male Yorkshire pigs (30-40 kg) were hemorrhaged to a mean arterial pressure (MAP) of 35 to 40 mm Hg. Once lactate reached 7 mmol/L, either saline (n = 5) or 10% PEG-20k (n = 5) was rapidly infused at 10% calculated blood volume. The primary outcome was LVR time, defined by the time from LVR administration to the time when lactate again reached 7 mmol/L. Other outcomes measured included MAP, heart rate, cardiac output, mixed venous oxygen saturation, splanchnic blood flow, and hemoglobin. Relative to saline, PEG-20k given after controlled hemorrhage increased LVR time by 16-fold, a conservative estimate given that the lactate never rose after LVR in the PEG-20k group. Survival was 80% for PEG-20k LVR compared to 0% for the saline controls (p the intravascular compartment. In a preclinical model of controlled hemorrhagic shock, PEG-20k-based LVR solution increased tolerance to the shock state 16-fold compared to saline. Polyethylene glycol-20k is a superior crystalloid for LVR that may increase safe transport times in the prehospital setting and find use in hospital emergency departments and operating rooms for patients awaiting volume replacement or normalization of cell, tissue, and compartment fluid volumes.

Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

Energy Technology Data Exchange (ETDEWEB)

Lai, W.; McCauley, E.W.

1978-01-04

Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Lai, W.; McCauley, E.W.

1978-01-01

Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

Electrochemical oxidation of propene by use of LSM₁₅/CGO₁₀ electrochemical reactor

DEFF Research Database (Denmark)

Ippolito, Davide; Andersen, Kjeld Bøhm; Kammer Hansen, Kent

2012-01-01

The propene catalytic oxidation was studied over an 11-layers porous electrochemical reactor made by La0.85Sr0.15MnO3 and Ce0.9Gd0.1O1.95 with the objective to simulate the abatement of exhaust gases emitted from Diesel engines. This work shows the possibility to enhance the catalytic activity th...... of catalysis (EPOC) was found at low temperature....

Reduced metabolic disease risk profile by voluntary wheel running accompanying juvenile Western diet in rats bred for high and low voluntary exercise.

Science.gov (United States)

Ruegsegger, Gregory N; Toedebusch, Ryan G; Braselton, Joshua F; Roberts, Christian K; Booth, Frank W

2015-12-01

Metabolic disease risk is influenced by genetics and modifiable factors, such as physical activity and diet. Beginning at 6 weeks of age, rats selectively bred for high (HVR) versus low voluntary running distance (LVR) behaviors were housed in a complex design with or without voluntary running wheels being fed either a standard or Western (WD, 42% kcal from fat and added sucrose) diet for 8 weeks. Upon intervention completion, percent body fat, leptin, insulin, and mediobasal hypothalamic mRNAs related to appetite control were assessed. Wheel access led to differences in body weight, food intake, and serum leptin and insulin. Intriguingly, percent body fat, leptin, and insulin did not differ between HVR and LVR lines in response to the two levels of voluntary running, regardless of diet, after the 8 wk. experiment despite HVR eating more calories than LVR regardless of diet and voluntarily running 5-7 times further in wheels than LVR. In response to WD, we observed increases in Cart and Lepr mediobasal hypothalamic mRNA in HVR, but no differences in LVR. Npy mRNA was intrinsically greater in LVR than HVR, while wheel access led to greater Pomc and Cart mRNA in LVR versus HVR. These data suggest that despite greater consumption of WD, HVR animals respond similarly to WD as LVR as a result, in part, of their increased wheel running behavior. Furthermore, high physical activity in HVR may offset the deleterious effects of a WD on adiposity despite greater energy intake in this group. Copyright © 2015 Elsevier Inc. All rights reserved.

Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 15, 2011, 3:30 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 15 mars 2011 a 15h30

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-07-01

This situation note is established according to the information gained on March 15, 2011, at 3:30 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 and of the spent fuel pools of reactors No. 5 and 6 of the Fukushima I site (Dai-ichi), 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.)

Physical parameters and biological effects of the LVR-15 epithermal neutron beam

Czech Academy of Sciences Publication Activity Database

Burian, J.; Gambarini, G.; Marek, M.; Mareš, Vladislav; Rejchrt, J.; Vanossi, E.; Viererbl, L.; Judas, L.

-, - (2006), s. 481-484 [International Congress on Neutron Capture Therapy /12./. Kagawo, 09.10.2006-13.10.2006] R&D Projects: GA MPO(CZ) 1H-PK2/05 Institutional research plan: CEZ:AV0Z50110509 Keywords : brain tumors * neutron capture therapy * NTC * dosimetry Subject RIV: FD - Oncology ; Hematology

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Fusion reactor problems

International Nuclear Information System (INIS)

Carruthers, R.

It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

RA research reactor - properties and experimental capabilities

International Nuclear Information System (INIS)

Milosevic, M.; Martinc, R.

1978-01-01

The brief survey of the Reactor RA exploitation experience, as well as the reactor equipment state, after 18 years of operation is presented. The results of efforts spent on reactor characteristics improvement in order to ensure safe and reliable reactor operation for next 15-20 years, are described [sr

Feasibility of Lung Volume Recruitment in Early Neuromuscular Weakness: A Comparison Between Amyotrophic Lateral Sclerosis, Myotonic Dystrophy, and Postpolio Syndrome.

Science.gov (United States)

Kaminska, Marta; Browman, Franceen; Trojan, Daria A; Genge, Angela; Benedetti, Andrea; Petrof, Basil J

2015-07-01

Lung volume recruitment (LVR) is a cough assistance technique used in persons with neuromuscular disorders (NMDs), most typically in those requiring noninvasive ventilation (NIV). Whether it may be useful in persons with NMDs who have milder respiratory impairment is unknown. To assess the feasibility, impact on quality of life (QOL), and preliminary physiological effects of daily LVR in different categories of persons with NMDs who have an early stage of respiratory impairment. Feasibility study. Academic tertiary care center. Outpatients diagnosed with amyotrophic lateral sclerosis (n = 8), postpolio syndrome (n = 10), and myotonic dystrophy (n = 6) who had restrictive respiratory defects but were not yet using NIV. Participants were asked to perform LVR up to 4 times daily and log their LVR use in a diary. Physiological measurements and questionnaires were completed at baseline and after 3 months. Compliance with LVR use was assessed, along with QOL and willingness to continue the treatment. Physiological measurements included forced vital capacity (FVC), lung insufflation capacity (LIC), and the LIC minus FVC difference. Of the 24 recruited subjects, 7 with amyotrophic lateral sclerosis, 7 with postpolio syndrome, and 5 with myotonic dystrophy completed the study (n = 19). At baseline, mean values for FVC and spontaneous peak cough flow were 59.9% predicted and 373.1 L/min, respectively. For subjects completing the study, 74% were willing to continue long-term LVR use, and QOL scores were not adversely affected by LVR in any NMD subgroup. The LIC-FVC difference increased from baseline to follow-up by a mean of 0.243 L (P = .006) in all subjects (n = 19), suggesting a possible improvement in respiratory system mechanics. In patients with NMDs who have early restrictive respiratory defects but do not yet require NIV, regular use of LVR is feasible with no negative impact on QOL over a 3-month period and may have physiological benefits. Further work is needed to

The SPHINX reactor for engineering tests

International Nuclear Information System (INIS)

Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

1990-01-01

A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

Conference Proposal for the Organization of ISCRE 15 (summary of results)

International Nuclear Information System (INIS)

Tsotsis, Theodore

2000-01-01

Though ISCRE 15 goal was to emphasize the fundamentals, of equal importance were the technological advances currently shaping the future of the field. Topical areas for ISCRE 15 included: (1) Reactors for Materials Processing; (2) Waste Minimization and Remediation; (3) Environmentally Benign Processing; (4) Reactor Dynamics; (5) Reactor Control and Safety; (6) Reactor Scale-up and Economic Evaluation; (7) Computational and Modeling Aspects of Reaction/ Reactor Engineering; (8)Fluid-Solid Catalytic and Non-Catalytic Reaction Systems; and (9) Catalytic, Polymerization and Biochemical Reactors

Reactor Neutron Sources

International Nuclear Information System (INIS)

Aksenov, V.L.

1994-01-01

The present status and the prospects for development of reactor neutron sources for neutron scattering research in the world are considered. The fields of application of neutron scattering relative to synchrotron radiation, the creation stages of reactors (steady state and pulsed) and their position in comparison with spallation neutron sources at present and in the foreseen future are discussed. (author). 15 refs.; 8 figs.; 3 tabs

Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Dixon, D.; Lord, D.

1978-01-01

The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs

Comparison of Cobalt based Catalysts Supported on MWCNT and SBA-15 Supporters for Fischer-tropsch Synthesis by Using Novel Vortex Type Reactor

International Nuclear Information System (INIS)

Yakubov, A.; Shahrun, M.S.; Kutty, M.G.; Hamid, S.B.A.; Piven, V.

2011-01-01

10 and 40 wt% Co/ Multi wall Carbon Nano tubes (MWCNT) and 10 and 40 wt% Co/ Santa Barbara Amorphous-15 (SBA) catalysts were prepared via incipient wetness impregnation and characterized by Scanning Electron Microscopy equipped with Energy Dispersive X-ray Spectroscopy (SEM and EDX), N 2 adsorption-desorption (BET), X-ray Diffractometry (XRD), Transmission Electron Microscopy (TEM) and Temperature- Programmed Reduction and H 2 desorption TPD/RO. Co(NO 3 ) 2 * 6H 2 O was used as a cobalt precursor. 200 ml hastelloy autoclave reactor was implemented to see the performance of the catalysts. This report presents details about the catalyst synthesis and reactor study. (author)

The national standards program for research reactors

International Nuclear Information System (INIS)

Whittemore, W.L.

1977-01-01

In 1970 a standards committee called ANS-15 was established by the American Nuclear Society (ANS) to prepare appropriate standards for research reactors. In addition, ANS acts as Secretariat for a national standards committee N17 which is responsible to the American National Standards Institute (ANSI) for the national consensus efforts for standards related to research reactors. To date ANS-15 has completed or is working on 14 standards covering all aspects of the operation of research reactors. Of the 11 research reactor standards submitted to the ANSI N17 Committee since its inception, six have been issued as National standards, and the remaining are still in the process of review. (author)

Lung volume recruitment acutely increases respiratory system compliance in individuals with severe respiratory muscle weakness

Directory of Open Access Journals (Sweden)

Yannick Molgat-Seon

2017-03-01

Full Text Available The aim of the present study was to determine whether lung volume recruitment (LVR acutely increases respiratory system compliance (Crs in individuals with severe respiratory muscle weakness (RMW. Individuals with RMW resulting from neuromuscular disease or quadriplegia (n=12 and healthy controls (n=12 underwent pulmonary function testing and the measurement of Crs at baseline, immediately after, 1 h after and 2 h after a single standardised session of LVR. The LVR session involved 10 consecutive supramaximal lung inflations with a manual resuscitation bag to the highest tolerable mouth pressure or a maximum of 50 cmH2O. Each LVR inflation was followed by brief breath-hold and a maximal expiration to residual volume. At baseline, individuals with RMW had lower Crs than controls (37±5 cmH2O versus 109±10 mL·cmH2O−1, p0.05. LVR had no significant effect on measures of pulmonary function at any time point in either group (all p>0.05. During inflations, mean arterial pressure decreased significantly relative to baseline by 10.4±2.8 mmHg and 17.3±3.0 mmHg in individuals with RMW and controls, respectively (both p<0.05. LVR acutely increases Crs in individuals with RMW. However, the high airway pressures during inflations cause reductions in mean arterial pressure that should be considered when applying this technique.

International Working Group on Fast Reactors Eight Annual Meeting, Vienna, Austria, 15-18 April 1975. Summary Report. Part II

International Nuclear Information System (INIS)

1975-07-01

The Eighth Annual Meeting of the IAEA International Working Group on Past Reactors was held at the IAEA Headquarters in Vienna, Austria, from 15 to 18 April 1975. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programmes in the field of LMPBR’s and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

Specialists' meeting on gas-cooled reactor core and high temperature instrumentation, Windermere, UK, 15-17 June 1982. Summary report

International Nuclear Information System (INIS)

1982-09-01

The Specialists' Meeting on ''Gas-Cooled Reactor Core and High Temperature Instrumentation'' was held at the Beech Hill Hotel, Windermere in England on June 15-17 1982. The meeting was sponsored by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors and was hosted by the Windscale Nuclear Power Development Laboratories of the UKAEA. The meeting was attended by 43 participants from Belgium, France, Federal Republic of Germany, Japan, United Kingdom of Great Britain and Northern Ireland and the United States of America. The objective of the meeting was to provide a forum, both formal and informal, for the exchange and discussion of technical information relating to instrumentation being used or under development for the measurement of core parameters, neutron flux, temperature, coolant flow etc. in gas cooled reactors. The technical part of the meeting was divided into five subject sessions: (A) Temperature Measurement (B) Neutron Detection Instrumentation (C) HTR Instrumentation - General (D) Gas Analysis and Failed Fuel Detection (E) Coolant Mass Flow and Leak Detection. A total of twenty-five papers were presented by the participants on behalf of their organizations during the meeting. A programme of the meeting and list of participants are given in appendices to this report

Lung Volume Reduction Coil Treatment in Chronic Obstructive Pulmonary Disease Patients with Homogeneous Emphysema : A Prospective Feasibility Trial

NARCIS (Netherlands)

Klooster, Karin; ten Hacken, Nick H. T.; Franz, Ina; Kerstjens, Huib A. M.; van Rikxoort, Eva M.; Slebos, Dirk-Jan

2014-01-01

Background: In patients with heterogeneous emphysema, surgical and bronchoscopic lung volume reduction (LVR) treatments are available. However, for patients with homogeneous emphysema these treatments are hardly investigated and seem less effective. Bronchoscopic LVR coil treatment has been shown to

Transmutation of Tc-99 in fission reactors

International Nuclear Information System (INIS)

Kloosterman, J.L.; Li, J.M.

1994-12-01

Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

Probabilistic record linkage is a valid and transparent tool to combine databases without a patient identification number

NARCIS (Netherlands)

Meray, Nora; Reitsma, Johannes B.; Ravelli, Anita C. J.; Bonsel, Gouke J.

2007-01-01

OBJECTIVE: To describe the technical approach and subsequent validation of the probabilistic linkage of the three anonymous, population-based Dutch Perinatal Registries (LVR1 of midwives, LVR2 of obstetricians, and LNR of pediatricians/neonatologists). These registries do not share a unique

Lung volume reduction coil treatment for patients with severe emphysema : a European multicentre trial

NARCIS (Netherlands)

Deslee, Gaetan; Klooster, Karin; Hetzel, Martin; Stanzel, Franz; Kessler, Romain; Marquette, Charles-Hugo; Witt, Christian; Blaas, Stefan; Gesierich, Wolfgang; Herth, Felix J. F.; Hetzel, Juergen; van Rikxoort, Eva M.; Slebos, Dirk-Jan

2014-01-01

Background The lung volume reduction (LVR) coil is a minimally invasive bronchoscopic nitinol device designed to reduce hyperinflation and improve elastic recoil in severe emphysema. We investigated the feasibility, safety and efficacy of LVR coil treatment in a prospective multicentre cohort trial

Cocaine-induced locomotor activity in rats selectively bred for low and high voluntary running behavior.

Science.gov (United States)

Brown, Jacob D; Green, Caroline L; Arthur, Ian M; Booth, Frank W; Miller, Dennis K

2015-02-01

The rewarding effects of physical activity and abused drugs are caused by stimulation of similar brain pathways. Low (LVR) and high (HVR) voluntary running lines were developed by selectively breeding Wistar rats on running distance performance on postnatal days 28-34. We hypothesized that LVR rats would be more sensitive to the locomotor-activating effects of cocaine than HVR rats due to their lower motivation for wheel running. We investigated how selection for LVR or HVR behavior affects inherited activity responses: (a) open field activity levels, (b) habituation to an open field environment, and (c) the locomotor response to cocaine. Open field activity was measured for 80 min on three successive days (days 1-3). Data from the first 20 min were analyzed to determine novelty-induced locomotor activity (day 1) and the habituation to the environment (days 1-3). On day 3, rats were acclimated to the chamber for 20 min and then received saline or cocaine (10, 20, or 30 mg/kg) injection. Dopamine transporter (DAT) protein in the nucleus accumbens was measured via Western blot. Selecting for low and high voluntary running behavior co-selects for differences in inherent (HVR > LVR) and cocaine-induced (LVR > HVR) locomotor activity levels. The differences in the selected behavioral measures do not appear to correlate with DAT protein levels. LVR and HVR rats are an intriguing physical activity model for studying the interactions between genes related to the motivation to run, to use drugs of abuse, and to exhibit locomotor activity.

Biologic lung volume reduction in advanced upper lobe emphysema: phase 2 results.

Science.gov (United States)

Criner, Gerard J; Pinto-Plata, Victor; Strange, Charlie; Dransfield, Mark; Gotfried, Mark; Leeds, William; McLennan, Geoffrey; Refaely, Yael; Tewari, Sanjiv; Krasna, Mark; Celli, Bartolome

2009-05-01

Biologic lung volume reduction (BioLVR) is a new endobronchial treatment for advanced emphysema that reduces lung volume through tissue remodeling. Assess the safety and therapeutic dose of BioLVR hydrogel in upper lobe predominant emphysema. Open-labeled, multicenter phase 2 dose-ranging studies were performed with BioLVR hydrogel administered to eight subsegmental sites (four in each upper lobe) involving: (1) low-dose treatment (n = 28) with 10 ml per site (LD); and (2) high-dose treatment (n = 22) with 20 ml per site (HD). Safety was assessed by the incidence of serious medical complications. Efficacy was assessed by change from baseline in pulmonary function tests, dyspnea score, 6-minute walk distance, and health-related quality of life. After treatment there were no deaths and four serious treatment-related complications. A reduction in residual volume to TLC ratio at 12 weeks (primary efficacy outcome) was achieved with both LD (-6.4 +/- 9.3%; P = 0.002) and HD (-5.5 +/- 9.4%; P = 0.028) treatments. Improvements in pulmonary function in HD (6 mo: DeltaFEV(1) = +15.6%; P = 0.002; DeltaFVC = +9.1%; P = 0.034) were greater than in LD patients (6 mo: DeltaFEV(1) = +6.7%; P = 0.021; DeltaFVC = +5.1%; P = 0.139). LD- and HD-treated groups both demonstrated improved symptom scores and health-related quality of life. BioLVR improves physiology and functional outcomes up to 6 months with an acceptable safety profile in upper lobe predominant emphysema. Overall improvement was greater and responses more durable with 20 ml per site than 10 ml per site dosing. Clinical trial registered with www.clinicaltrials.gov (NCT 00435253 and NCT 00515164).

Bronchoscopic Lung Volume Reduction Coil Treatment of Patients With Severe Heterogeneous Emphysema

NARCIS (Netherlands)

Slebos, Dirk-Jan; Klooster, Karin; Ernst, Armin; Herth, Felix J. F.; Kerstjens, Huib A. M.

Background: The lung volume reduction coil (LVR-coil), a new experimental device to achieve lung volume reduction by bronchoscopy in patients with severe emphysema, works in a manner unaffected by collateral airflow. We investigated the safety and efficacy of LVR-coil treatment in patients with

Correlations between power and test reactor data bases

International Nuclear Information System (INIS)

Guthrie, G.L.; Simonen, E.P.

1989-02-01

Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 15, 2011, 10:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011 (15 mars 2011 - point a 22h00 heures)

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-07-01

This situation note is established according to the information gained on March 15, 2011, at 10:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 and of the spent fuel pools of reactors No. 4, 5 and 6 of the Fukushima I site (Dai-ichi), 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.)

Directions for improved fusion reactors

International Nuclear Information System (INIS)

Krakowski, R.A.; Miller, R.L.; Delene, J.G.

1986-01-01

Conceptual fusion reactor studies over the past 10 to 15 years have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points towards smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. A generic fusion physics/engineering/costing model is used to provide a quantiative basis for these arguments for specific fusion concepts

RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

1981-12-01

The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

Nuclear reactor monitoring device

International Nuclear Information System (INIS)

Mihashi, Ishi; Honma, Hitoshi.

1993-01-01

The monitoring device of the present invention comprises a reactor core/reactor system data measuring and controlling device, a radioactivity concentration calculation device for activated coolants for calculating a radioactivity concentration of activated coolants in a main steam and reactor water by using an appropriate physical model, a radioactivity concentration correlation and comparison device for activated coolants for comparing correlationship with a radiation dose and an abnormality alarm device. Since radioactivity of activated primary coolants is monitored at each of positions in the reactor system and occurrence of leakage and the amount thereof from a primary circuit to a secondary circuit is monitored if the reactor has secondary circuit, integrity of the reactor system can be ensured and an abnormality can be detected rapidly. Further, radioactivity concentration of activated primary circuit coolants, represented by 16 N or 15 C, is always monitored at each of positions of PWR primary circuits. When a heat transfer pipe is ruptured in a steam generator, leakage of primary circuit coolants is detected rapidly, as well as the amount of the leakage can be informed. (N.H.)

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

Radiation damage in a high Ni weld

International Nuclear Information System (INIS)

Brumovsky, M.; Kytka, M.; Kopriva, R.

2015-01-01

WWER-1000 RPV weld metals are characterized by a high content of nickel, mostly about 1.7 mass % with content of manganese around 0.8 mass % with a very low copper content - about 0.05 mass %. In such material some late blooming phase effect should be observed during irradiation. Such typical weld material was irradiated in the experimental reactor LVR-15 in N RI Rez at the irradiation temperature 290 C degrees and at five neutron fluences from 1.5 to 9.5 *10 23 m -2 (E>1 MeV). Charpy V-notch impact tests, static fracture toughness tests, tensile and hardness measurement were performed to obtain effect of neutron fluence on radiation hardening as well as embrittlement. Neutron fluence dependences of all these property changes have monotonic character but with a high neutron embrittlement exponent around 0.8. Scanning electron microscope of fracture surfaces showed no or very small portion of intercrystalline fracture. Transmission electron microscopy was performed on specimens from all neutron fluences. Only low density of black-dot damage has been observed. It is assumed that most of defect are dislocation loops. The late blooming phase which may be observed from results of mechanical properties are probably below the resolution of the used JEM-2010, i.e. 1.5 nm. (authors)

Mu-opioid receptor inhibition decreases voluntary wheel running in a dopamine-dependent manner in rats bred for high voluntary running.

Science.gov (United States)

Ruegsegger, Gregory N; Brown, Jacob D; Kovarik, M Cathleen; Miller, Dennis K; Booth, Frank W

2016-12-17

The mesolimbic dopamine and opioid systems are postulated to influence the central control of physical activity motivation. We utilized selectively bred rats for high (HVR) or low (LVR) voluntary running behavior to examine (1) inherent differences in mu-opioid receptor (Oprm1) expression and function in the nucleus accumbens (NAc), (2) if dopamine-related mRNAs, wheel-running, and food intake are differently influenced by intraperitoneal (i.p.) naltrexone injection in HVR and LVR rats, and (3) if dopamine is required for naltrexone-induced changes in running and feeding behavior in HVR rats. Oprm1 mRNA and protein expression were greater in the NAc of HVR rats, and application of the Oprm1 agonist [D-Ala2, N-MePhe4, Gly-ol]-enkephalin (DAMGO) to dissociated NAc neurons produced greater depolarizing responses in neurons from HVR versus LVR rats. Naltrexone injection dose-dependently decreased wheel running and food intake in HVR, but not LVR, rats. Naltrexone (20mg/kg) decreased tyrosine hydroxylase mRNA in the ventral tegmental area and Fos and Drd5 mRNA in NAc shell of HVR, but not LVR, rats. Additionally, lesion of dopaminergic neurons in the NAc with 6-hydroxydopamine (6-OHDA) ablated the decrease in running, but not food intake, in HVR rats following i.p. naltrexone administration. Collectively, these data suggest the higher levels of running observed in HVR rats, compared to LVR rats, are mediated, in part, by increased mesolimbic opioidergic signaling that requires downstream dopaminergic activity to influence voluntary running, but not food intake. Copyright © 2016 IBRO. Published by Elsevier Ltd. All rights reserved.

Functional Outcomes of the Low Vision Depression Prevention Trial in Age-Related Macular Degeneration.

Science.gov (United States)

Deemer, Ashley D; Massof, Robert W; Rovner, Barry W; Casten, Robin J; Piersol, Catherine V

2017-03-01

To compare the efficacy of behavioral activation (BA) plus low vision rehabilitation with an occupational therapist (OT-LVR) with supportive therapy (ST) on visual function in patients with age-related macular degeneration (AMD). Single-masked, attention-controlled, randomized clinical trial with AMD patients with subsyndromal depressive symptoms (n = 188). All subjects had two outpatient low vision rehabilitation optometry visits, then were randomized to in-home BA + OT-LVR or ST. Behavioral activation is a structured behavioral treatment aiming to increase adaptive behaviors and achieve valued goals. Supportive therapy is a nondirective, psychological treatment that provides emotional support and controls for attention. Functional vision was assessed with the activity inventory (AI) in which participants rate the difficulty level of goals and corresponding tasks. Participants were assessed at baseline and 4 months. Improvements in functional vision measures were seen in both the BA + OT-LVR and ST groups at the goal level (d = 0.71; d = 0.56 respectively). At the task level, BA + OT-LVR patients showed more improvement in reading, inside-the-home tasks and outside-the-home tasks, when compared to ST patients. The greatest effects were seen in the BA + OT-LVR group in subjects with a visual acuity ≥20/70 (d = 0.360 reading; d = 0.500 inside the home; d = 0.468 outside the home). Based on the trends of the AI data, we suggest that BA + OT-LVR services, provided by an OT in the patient's home following conventional low vision optometry services, are more effective than conventional optometric low vision services alone for those with mild visual impairment. (ClinicalTrials.gov number, NCT00769015.).

Radiation effects on reactor pressure vessel supports

International Nuclear Information System (INIS)

Johnson, R.E.

1996-05-01

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue

What occurred in the reactors

International Nuclear Information System (INIS)

Kudo, Kazuhiko

2013-01-01

Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

Accident at the zero power reactor which happened on October 15 1958

International Nuclear Information System (INIS)

Savic, P.

1959-01-01

During an experiment on the zero power heavy water reactor with natural uranium fuel in the Boris Kidric Institute of Nuclear Sciences, the reactor escaped control. Six staff members in the immediate surrounding of the bare assembly were exposed to high neutron and ionising irradiation. Other two employees who were at some bigger distance were exposed to doses higher than permitted. This paper deals with the circumstances that caused the accident, status of the dosimetry, control and alarm systems. Individual exposure doses were estimated according to the calculated neutron flux values obtained from measuring the activities of personal belongings made of gold and copper as well as radioactive phosphorous from urine

CAREM project 15 to 150 MWe

International Nuclear Information System (INIS)

1992-05-01

The main goal of the Carem Project is the introduction of a Inherent Safe Nuclear Power Reactor in the range of low power (15 to 150 MWe). For this low-power application, light-water and low enriched uranium was selected, since using those concepts permits to take full advantage of the special characteristics of low power reactors. INVAP has been involved in the last years in the design and construction of a Carem Reactor, which could cover a range up to 150 MWe, using a multiple-unit approach. It would furnished the 150 MWe, using six Carem Reactors, of proper power, which would share most of the services. INVAP is a reliable supplier of not only the nuclear reactor but also of the fuel

Nucleus accumbens neuronal maturation differences in young rats bred for low versus high voluntary running behaviour

Science.gov (United States)

Roberts, Michael D; Toedebusch, Ryan G; Wells, Kevin D; Company, Joseph M; Brown, Jacob D; Cruthirds, Clayton L; Heese, Alexander J; Zhu, Conan; Rottinghaus, George E; Childs, Thomas E; Booth, Frank W

2014-01-01

We compared the nucleus accumbens (NAc) transcriptomes of generation 8 (G8), 34-day-old rats selectively bred for low (LVR) versus high voluntary running (HVR) behaviours in rats that never ran (LVRnon-run and HVRnon-run), as well as in rats after 6 days of voluntary wheel running (LVRrun and HVRrun). In addition, the NAc transcriptome of wild-type Wistar rats was compared. The purpose of this transcriptomics approach was to generate testable hypotheses as to possible NAc features that may be contributing to running motivation differences between lines. Ingenuity Pathway Analysis and Gene Ontology analyses suggested that ‘cell cycle’-related transcripts and the running-induced plasticity of dopamine-related transcripts were lower in LVR versus HVR rats. From these data, a hypothesis was generated that LVR rats might have less NAc neuron maturation than HVR rats. Follow-up immunohistochemistry in G9–10 LVRnon-run rats suggested that the LVR line inherently possessed fewer mature medium spiny (Darpp-32-positive) neurons (P running wheel access in our G9–10 LVRs uniquely increased their Darpp-32-positive and Dcx-positive neuron densities. In summary, NAc cellularity differences and/or the lack of running-induced plasticity in dopamine signalling-related transcripts may contribute to low voluntary running motivation in LVR rats. PMID:24665095

Optimal time for predicting left ventricular remodeling after successful primary coronary angioplasty in acute myocardial infarction using serial myocardial contrast echocardiography and magnetic resonance imaging

International Nuclear Information System (INIS)

Sakuma, Tadamichi; Okada, Takenori; Hayashi, Yasuhiko; Otsuka, Masaya; Hirai, Yuukou

2002-01-01

The objective of this study was to determine the optimal time to assess microvascular integrity within the risk area for myocardial infarction in order to predict unfavorable left ventricular remodeling (LVR) after successful primary coronary angioplasty. Fifty-three patients who underwent myocardial contrast echocardiography (MCE) just before recanalization, shortly after and 1 day (Day 2) and 3 weeks after recanalization were studied. The no- and low-reflow ratio (LR ratio) was analyzed at each stage. The wall-tinning ratio within the risk area was determined using magnetic resonance imaging performed 3-4 weeks after the recanalization. Thirteen of the 53 patients showed LVR 3-8 months after recanalization. The optimal time to predict LVR was found to be Day 2 based on the receiver operating characteristic curves. The LR ratio on Day 2 (χ 2 =7.39, p=0.007) and the collateral circulation before recanalization (χ 2 =4.57, p=0.03) were chosen as independent variables for predicting LVR. Patients with greater than 0.43 in the LR ratio on Day 2 showed a lower wall-thinning ratio (58±19% vs 72±20%, p=0.05). This study shows that the optimal time to estimate the microvascular integrity for predicting LVR is 1 day after recanalization, which is neither shortly after recanalization nor during the convalescent stage. (author)

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

The status of the PIK reactor

Energy Technology Data Exchange (ETDEWEB)

Petrov, Yu V [Academy of Sciences of Russia, Petersburg Nuclear Physics Institute, Gatchina, St. Petersburg (Russian Federation)

1992-07-01

This report describes the 100 MW research reactor PIK which is now under construction. The thermal neutron flux in the heavy water reflector exceeds 10{sup 15} cm{sup -2}s{sup -1}; in the light water trap, it is about 4{center_dot}10{sup 15} cm{sup -2}s{sup -1}. The replaceable core vessel allows to vary the parameters of the core over a wide range. The reactor provides sources of hot, cold and ultracold neutrons for 10 horizontal, 6 inclined neutron beams, and 8 neutron guides. At the ends of the beam tubes, the neutron flux is 10{sup 10} - 10{sup 11} cm{sup -2}s{sup -1}. The flux of the long wave neutrons exceeds 10{sup 9} cm{sup -2}s{sup -1}. To ensure precise measurements, the experimental hall is protected against vibrations. The project meets all modern safety requirements. The calculated parameters of the reactor were verified using a full-scale mock-up. Seventy percent of the reactor construction and installation were completed in the beginning of 1992. (author)

Transmutation of actinide 237Np with a fusion reactor and a hybrid reactor

International Nuclear Information System (INIS)

Feng, K.M.; Huang, J.H.

1994-01-01

The use of fusion reactors to transmute fission reactor wastes to stable species is an attractive concept. In this paper, the feasibility of transmutation of the long-lived actinide radioactive waste Np-237 with a fusion reactor and a hybrid reactor has been investigated. A new waste management concept of burning HLW (High Level Waste), utilizing released energy and converting Np-237 into fissile fuel Pu-239 through transmutation has been adopted. The detailed neutronics and depletion calculation of waste inventories was carried out with a modified version of one-dimensional neutron transport and burnup calculation code system BISON1.5 in this study. The transmutation rate of Np with relationship to neutron wall loading, Pu and Np with relationship to neutron wall load, Pu and Np concentration in the transmutation zone have been explored as well as relevant results are also given

Compact toroid refueling of reactors

International Nuclear Information System (INIS)

Gouge, M.J.; Hogan, J.T.; Milora, S.L.; Thomas, C.E.

1988-04-01

The feasibility of refueling fusion reactors and devices such as the International Thermonuclear Engineering Reactor (ITER) with high-velocity compact toroids is investigated. For reactors with reasonable limits on recirculating power, it is concluded that the concept is not economically feasible. For typical ITER designs, the compact toroid fueling requires about 15 MW of electrical power, with about 5 MW of thermal power deposited in the plasma. At these power levels, ideal ignition (Q = ∞) is not possible, even for short-pulse burns. The pulsed power requirements for this technology are substantial. 6 ref., 1 figs

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Mock-up experiments for the project of high dose irradiation on the RPV concrete

International Nuclear Information System (INIS)

Zdarek, J.; Brabec, P.; Frybort, O.; Lahodova, Z.; Vit, J.; Stemberk, P.

2015-01-01

Aging of NPP's concrete structures comes into growing interest in connection with solution of life extension programmes of operated units. Securing continued safe operation of NPPs calls for additional proofs of suitable long term behaviour of loaded reinforced concrete structures. An irradiation test of concrete samples was performed in the core of the LVR-15 reactor. The irradiation capsule was hung in the irradiation channel and the cooling of the capsule was ensured through direct contact of the capsule wall with the primary circuit water. Cylindrical, serpentine concrete samples (50 mm in diameter and 100 mm in length), representing composition of WWER RPV cavity, was chosen as a compromise of mechanical properties testing needs and dimension limitations of reactor irradiation channel. Heating during irradiation test was maintained under 93 Celsius degrees by cooling and was controlled by embedded thermocouple. Design of the cooling management was supported by computational analysis. The dependencies of heated concrete samples to the neutron fluence and the gamma heating were obtained by changing the thermal power of the reactor and by changing the vertical position of the sample in the irradiation channel. The irradiation capsule was filled with inert gas (helium) to allow the measurement of generated gas. The determination of concrete samples activity for long-term irradiation was performed on the principles of the Neutron Activation Analysis. Preliminary mock-up tests have proved the ability to fulfill technical needs for planned high dose irradiation experiment

Thermal-hydraulic analyses for in-pile SCWR fuel qualification test loops and SCWR material loop

Energy Technology Data Exchange (ETDEWEB)

Vojacek, A.; Mazzini, G.; Zmitkova, J.; Ruzickova, M. [Research Centre Rez (Czech Republic)

2014-07-01

One of the R&D directions of Research Centre Rez is dedicated to the supercritical water-cooled reactor concept (SCWR). Among the developed experimental facilities and infrastructure in the framework of the SUSEN project (SUStainable ENergy) is construction and experimental operation of the supercritical water loop SCWL focusing on material tests. At the first phase, this SCWL loop is assembled and operated out-of-pile in the dedicated loop facilities hall. At this out-of-pile operation various operational conditions are tested and verified. After that, in the second phase, the SCWL loop will be situated in-pile, in the core of the research reactor LVR-15, operated at CVR. Furthermore, it is planned to carry out a test of a small scale fuel assembly within the SuperCritical Water Reactor Fuel Qualification Test (SCWR-FQT) loop, which is now being designed. This paper presents the results of the thermal-hydraulic analyses of SCWL loop out-of-pile operation using the RELAP5/MOD3.3. The thermal-hydraulic modeling and the performed analyses are focused on the SCWL loop model validation through a comparison of the calculation results with the experimental results obtained at various operation conditions. Further, the present paper focuses on the transient analyses for start-up and shut-down of the FQT loop, particularly to explore the ability of system codes ATHLET 3.0A to simulate the transient between subcritical conditions and supercritical conditions. (author)

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Nuclear Reactor RA Safety Report, Vol. 15, Analysis of significant accidents; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 15, Analiza znacajnih akcidenata

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-11-01

Power excursions of the RA reactor a mathematical model of reactor kinetic behaviour was formulated to describe power and temperature coefficients for both reactor fuel and moderator. Computer code TM-1 was written for analysis of possible reactor accidents. Power excursions caused by uncontrolled control rod removal and heavy water flow into the central vertical experimental channel were analyzed. Accidents caused by fuel elements handling were discussed including possible fuel element damage. Although the probability for uncontrolled radioactive materials release into the environment is very low, this type of accidents are analyzed as well including the impact on the personnel and the environment. A separate chapter describes analysis of the loss of flow accident. Safety analysis covers the possible damage of the outer steel Ra reactor vessel and the water screens which are part of the water biological shield. [Serbo-Croat] Radi analize dinamickog ponasanja reaktora RA u toku ekskurzije snage formulisan je matematicki model koji opisuje promene snage i temperature goriva i moderatora. Za analizu predpostavljenih akcidenata realizovan je racunarski program TM-1. Analizirane su ekskurzije snage usled nekontrolisanog izvlacenja kontrolnih sipki i punjenja centralnog vertikalnog eksperimentalnog kanala teskom vodom. Analizirani su akcidenti pri rukovanju nuklearnim gorivom ukljucujuci ostecenje nuklearnog goriva. Iako je verovatnoca za nekontrolisano rasturanje radioaktivnog materijala mala, analizirani su i ovakvi moguci akcidenti koji mogu uticati kako na osoblje reaktora tako i na okolinu. Posebno poglavlje obuhvata analizu akcidenta u slucaju prestanka cirkulacije primarnog hladioca. Analiza sigurnosti reaktora obuhvata i moguce ostecenje spoljasnjeg celicnog suda reaktora RA kao i vodenih ekrana koji su deo vodenog bioloskog stita.

The Simulator Development for RDE Reactor

Science.gov (United States)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Ozone disintegration kinetics in the reactor for tyres decomposition

International Nuclear Information System (INIS)

Golota, V.I.; Manujlenko, O.V.; Taran, G.V.; Pis'menetskij, A.S.; Zamuriev, A.A.

2010-01-01

The results of theoretical and experimental research of ozone disintegration kinetics in the chemical reactor which is developed for decomposition of tyres in the ozone-air environment are presented. Analytical expression for dependence of ozone concentration in the reactor from time and from parameters of the task, such as volume speed of ozone-air mixture feed on a reactor input, concentration of ozone on the input to the reactor, volume speed of output of the used mixture, reactor size, and square of its internal surface is obtained. It is shown that at the same speed of ozone-air mixture pro rolling through the reactor, with growth of ozone concentration on the input, value of stationary concentration in the reactor grows, remaining always less than concentration on the input. It is also shown that at the same ozone concentration on the input, with growth of speed of ozone-air mixture pro rolling through the reactor, value of stationary ozone concentration in the reactor also grows, remaining always less than ozone concentration on the input. The ozone disintegration kinetics in the reactor in a wide range of speed of ozone-air mixture pro rolling through the reactor (0.15, 0.30, 0.45, 0.60 m3/hour) and various ozone concentration on the input (5, 10, 15, 20 g/m3) is experimentally studied. It is shown that experimental results with good accuracy coincide with the theoretical. Direct experiment showed the essential influence of the internal surface of the reactor on the ozone disintegration kinetics.

Effectiveness and efficacy of minimally invasive lung volume reduction surgery for emphysema.

Science.gov (United States)

Pertl, Daniela; Eisenmann, Alexander; Holzer, Ulrike; Renner, Anna-Theresa; Valipour, A

2014-01-01

Lung emphysema is a chronic, progressive and irreversible destruction of the lung tissue. Besides non-medical therapies and the well established medical treatment there are surgical and minimally invasive methods for lung volume reduction (LVR) to treat severe emphysema. This report deals with the effectiveness and cost-effectiveness of minimally invasive methods compared to other treatments for LVR in patients with lung emphysema. Furthermore, legal and ethical aspects are discussed. No clear benefit of minimally invasive methods compared to surgical methods can be demonstrated based on the identified and included evidence. In order to assess the different methods for LVR regarding their relative effectiveness and safety in patients with lung emphysema direct comparative studies are necessary.

Nuclear research reactor of Da Lat. Final report for the period 15 December 1987 - 14 December 1988

Energy Technology Data Exchange (ETDEWEB)

Long, Vu Hai [Viet Nam Atomic Energy Committee, Da Lat (Viet Nam). Dept. of Reactor Physics and Engineering

1991-12-31

This Safety Analysis Report aims at setting-up the balance of all the safety problems of the Dalat Nuclear Reactor with the standpoint and experience of 5 years operation and exploitation. It presents the characteristics of the site, the architecture and construction of the reactor, the design characteristics of the reactor core, control and instrumentation, radiation protection and environment radioactivity, safety analysis of abnormal situations, organization and administrative measures to ensure the safe operation and exploitation of the reactor. Refs, figs and tabs.

Nuclear research reactor of Da Lat. Final report for the period 15 December 1987 - 14 December 1988

International Nuclear Information System (INIS)

Vu Hai Long

1990-01-01

This Safety Analysis Report aims at setting-up the balance of all the safety problems of the Dalat Nuclear Reactor with the standpoint and experience of 5 years operation and exploitation. It presents the characteristics of the site, the architecture and construction of the reactor, the design characteristics of the reactor core, control and instrumentation, radiation protection and environment radioactivity, safety analysis of abnormal situations, organization and administrative measures to ensure the safe operation and exploitation of the reactor. Refs, figs and tabs

Archival of the ZPPR-15B physics experiment

International Nuclear Information System (INIS)

Lell, R.; McKnight, R.

2012-01-01

This I-NERI collaboration between Argonne National Laboratory (ANL) and Korea Atomic Energy Research Institute (KAERI) began mid-year (April, 2010). This report summarizes the progress for year two of the proposed three-year collaboration to generate a physics validation database of integral experiments for metallic fueled fast reactor systems. The objective of the proposed project is to archive and evaluate the integral experiment data, analyze the experiments, and prepare detailed computational models to be used for validating the modern suites of fast reactor design analysis tools which are under development at ANL and KAERI. A series of mockup experiments for a 330 MWe Integral Fast Reactor (IFR) at ANL under the ZPPR-15 Program, also known as the IFR Benchmark Physics Test Program will be retrieved and analyzed in this project. The ZPPR-15 program was conducted in four phases. Each phase was marked by a particular composition of the reference assembly. In the first phase (15A), only plutonium, depleted uranium, stainless steel and sodium were included in this very clean physics assembly. This allowed examination of the effect of removing oxygen from the typical oxide-fueled sodium fast reactor. Zirconium was added in the second phase (15B). Additionally, 13 control rods and channels were added after the first phase. In the third phase (15C), roughly half of the core volume was fueled by enriched uranium to simulate a fast reactor transition composition. In the final phase (15D), the enriched uranium component was increased to 90%, simulating a near-beginning-of-life composition. In addition to criticality, control rod worths, reaction rate distribution, reactivity coefficients, gamma heating, neutron spectrum and kinetics, there were a number of measurements aimed at addressing special issues of safety, economics and metal fuel composition. The BFS-73-1 and BFS-75-1 experiments of KAERI carried out as the mockup experiment of KALIMER-150 at the Russian BFS-1

Nonlinear dynamics of boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

1983-01-01

Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

Technology development and production of elongated shell for reactor vessel active zone of WWER-TOI project from steel 15Cr2NiMoVN class 1

International Nuclear Information System (INIS)

Shklyaev, S.Eh.; Titova, T.I.; Ratushev, D.V.; Shul'gan, N.A.; Eroshkin, S.B.; Durynin, V.A.; Efimov, S.V.; Dub, V.S.; Kulikov, A.P.; Romashkin, A.N.

2015-01-01

Production process for the elongated shell blank of the active zone of the reactor pressure vessel made from steel 15Cr2NiMoVN Class 1 with finished sizes Dext=4.655 mm, Dint=4.240 mm, H=4.910 mm (height for heat treatment – 5.750 mm) is presented. For the first time in Russia in production site of OMZ-Special steel LLC a unique elongated shell blank of the reactor vessel active zone was made from ingot 420.0 t for WWER-TOI project fully meeting the specified requirements in terms of metallurgical quality and set of service properties [ru

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

Durbin, M.E.; Montoya, G.M.

1991-06-01

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

Development of supercritical water reactors in Russia and abroad

International Nuclear Information System (INIS)

Glebov, A.P.; Klushin, A.V.

2014-01-01

The results of Russian and foreign studies on the water-cooled high critical parameters reactors are analyzed. Developments on this subject are conducted in more than 15 countries. The advantages of WWER- SCP and characteristics of experimental reactor of WWER-SCP-30 are discussed. It is noted that priority task is to develop a reactor with thermal neutron spectrum with a subsequent transition to the reactor with a fast neutron spectrum [ru

Very high flux steady state reactor and accelerator based sources

International Nuclear Information System (INIS)

Ludewig, H.; Todosow, M.; Simos, N.; Shapiro, S.; Hastings, J.

2004-01-01

With the number of steady state neutron sources in the US declining (including the demise of the Bnl HFBR) the remaining intense sources are now in Europe (i.e. reactors - ILL and FMR, accelerator - PSI). The intensity of the undisturbed thermal flux for sources currently in operation ranges from 10 14 n/cm 2 *s to 10 15 n/cm 2 *s. The proposed Advanced Neutron Source (ANS) was to be a high power reactor (about 350 MW) with a projected undisturbed thermal flux of 7*10 15 n/cm 2 *s but never materialized. The objective of the current study is to explore the requirements and implications of two source concepts with an undisturbed flux of 10 16 n/cm 2 *s. The first is a reactor based concept operating at high power density (10 MW/l - 15 MW/l) and a total power of 100 MW - 250 MW, depending on fissile enrichment. The second is an accelerator based concept relying on a 1 GeV - 1.5 GeV proton Linac with a total beam power of 40 MW and a liquid lead-bismuth eutectic target. In the reactor source study, the effects of fissile material enrichment, coolant temperature and pressure drop, and estimates of pressure vessel stress levels will be investigated. The fuel form for the reactor will be different from all other operating source reactors in that it is proposed to use an infiltrated graphitic structure, which has been developed for nuclear thermal propulsion reactor applications. In the accelerator based source the generation of spallation products and their activation levels, and the material damage sustained by the beam window will be investigated. (authors)

Dopamine D1 receptor modulation in nucleus accumbens lowers voluntary wheel running in rats bred to run high distances.

Science.gov (United States)

Roberts, Michael D; Gilpin, Leigh; Parker, Kyle E; Childs, Thomas E; Will, Matthew J; Booth, Frank W

2012-02-01

Dopamine signaling in the nucleus accumbens (NAc) has been postulated to influence reward development towards drugs of abuse and exercise. Herein, we used generation 4-5 rats that were selectively bred to voluntary run high (HVR) versus low (LVR) distances in order to examine if dopamine-like 1 (D1) receptor modulation in the NAc differentially affects nightly voluntary wheel running between these lines. A subset of generation 5-6 HVR and LVR rats were also used to study the mRNA expression of key genes related to reward and addiction in the NAc (i.e., DRD1, DRD5, DRD2, Nr4a2, FosB, and BDNF). In a crossover fashion, a D1-like agonist SKF 82958 (2 μg per side) or D1-like full antagonist SCH 23390 (4 μg per side) was bilaterally injected into the NAc of HVR and LVR female Wistar rats prior to their high running nights. Notably, during hours 2-4 (between 2000 and 2300) of the dark cycle there was a significant decrement in running distances in the HVR rats treated with the D1 agonist (p=0.025) and antagonist (p=0.017) whereas the running distances in LVR rats were not affected. Interestingly, HVR and LVR rats possessed similar NAc concentrations of the studied mRNAs. These data suggest that: a) animals predisposed to run high distances on a nightly basis may quickly develop a rewarding response to exercise due to an optimal D1-like receptor signaling pathway in the NAc that can be perturbed by either activation or blocking, b) D1-like agonist or antagonist injections do not increase running distances in rats that are bred to run low nightly distances, and c) running differences between HVR and LVR animals are seemingly not due to the expression of the studied mRNAs. Given the societal prevalence of obesity and extraneous physical inactivity, future studies should be performed in order to further determine the culprit for the low running phenotype observed in LVR animals. Copyright © 2011. Published by Elsevier Inc.

Backfitting swimming pool reactors

International Nuclear Information System (INIS)

Roebert, G.A.

1978-01-01

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

Today's attitudes and future prospects of fast reactors in Italy

International Nuclear Information System (INIS)

Barabaschi, S.; Cicognani, G.; Pierantoni, F.

1982-01-01

The Italian fast reactor programme is reviewed. The 15 year collaboration with France has resulted in the construction of the PEC reactor, development of the Superphenix-1 and a common R and D programme for future large fast reactors. The CNEN 4th five year (1980-84) plan is outlined. The budget breakdown for different areas shows the importance attached to the fast reactor. (U.K.)

Water treatment process for nuclear reactors

International Nuclear Information System (INIS)

Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

1993-01-01

Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigates water treatment process for nuclear reactor utilization. Analysis of outwater chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerants to obtain the optimum quantity of pure water which reached to 15 cubic-meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30%. Output water chemistry agrees with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined

Water treatment process for nuclear reactors

International Nuclear Information System (INIS)

Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

1992-01-01

Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigate water treatment process for nuclear reactor utilization. Analysis of output water chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerates to obtain the optimum quantity of pure water which reached to 15 cubic meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30 %. output water chemistry agree with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined.5 fig., 3 tab

NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

Energy Technology Data Exchange (ETDEWEB)

Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ugorowski, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ito, Takashi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

Development of the reactor safety film

International Nuclear Information System (INIS)

Sheheen, N.N.; Hodson, P.J.

1981-01-01

The first computer-generated film of LASL's Reactor Safety efforts was developed using the ANIMATE framework, a program that adds visual capabilities to MAPPER. Numerous software limitations had to be overcome within a very limited production schedule. A significant achievement was the 15,000-vector-per-frame sequence depicting a pressurized water reactor core with parts flashing while pumps circulate fluid through the system

Advanced gas cooled reactors - Designing for safety

International Nuclear Information System (INIS)

Keen, Barry A.

1990-01-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

Advanced gas cooled reactors - Designing for safety

Energy Technology Data Exchange (ETDEWEB)

Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

1990-07-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

Low Volume Resuscitation with Cell Impermeants

Science.gov (United States)

2016-04-01

of 10% Bovine Serum Albumin (BSA), a prototypical oncotic agent (n = 6). The outcomevariables for the study includedLVR time, plasma lactate, mean...re- quirement of bicarbonate administration to correct acidosis during resuscitation. The impermeant effect in LVR solutions is greatly aug- mented...resuscitation exacerbates TICS, acidosis , hypothermia, and coagulopathy (3, 4). Other resuscitation solutions such as hypertonic saline or starch have had

First preliminary design of an experimental fusion reactor

International Nuclear Information System (INIS)

1977-09-01

A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

Effectiveness and efficacy of minimally invasive lung volume reduction surgery for emphysema

Directory of Open Access Journals (Sweden)

Pertl, Daniela

2014-10-01

Full Text Available [english] Lung emphysema is a chronic, progressive and irreversible destruction of the lung tissue. Besides non-medical therapies and the well established medical treatment there are surgical and minimally invasive methods for lung volume reduction (LVR to treat severe emphysema. This report deals with the effectiveness and cost-effectiveness of minimally invasive methods compared to other treatments for LVR in patients with lung emphysema. Furthermore, legal and ethical aspects are discussed. No clear benefit of minimally invasive methods compared to surgical methods can be demonstrated based on the identified and included evidence. In order to assess the different methods for LVR regarding their relative effectiveness and safety in patients with lung emphysema direct comparative studies are necessary.

Backfitting of the FRG reactors

Energy Technology Data Exchange (ETDEWEB)

Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

1990-05-01

The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

Backfitting of the FRG reactors

International Nuclear Information System (INIS)

Krull, W.

1990-01-01

The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

On the optimization of a steady-state bootstrap-reactor

International Nuclear Information System (INIS)

Polevoy, A.R.; Martynov, A.A.; Medvedev, S.Yu.

1993-01-01

A commercial fusion tokamak-reactor may be economically acceptable only for low recirculating power fraction r 0 ≡ P CD /P α BS ≡I BS /I > 0.9 to sustain the steady-state operation mode for high plasma densities > 1.5 10 20 m -3 , fulfilled the divertor conditions. This paper presents the approximate expressions for the optimal set of reactor parameters for r BS /I∼1, based on the self-consistent plasma simulations by 1.5D ASTRA code. The linear MHD stability analysis for ideal n=1 kink and ballooning modes has been carried out to determine the conditions of stabilization for bootstrap steady state tokamak reactor BSSTR configurations. (author) 10 refs., 1 tab

Preliminary study or RSG-GAS reactor fuel element integrity

International Nuclear Information System (INIS)

Soejoedi, A.; Tarigan, A.; Sujalmo; Prayoga, S.; Suhadi

1996-01-01

After 8 years of operation, RSG-GAS was able to reach 15 cycles of reactor operation with 116 irradiated fuels, whereas 49 fuels were produced by NUKEM; and the other 67 were produced by PEBN-BATAN. At the 15 T h cycles, it have been used 40 standard fuels and 8 control fuels (Forty standard fuels and eight control fuels have been used in the 15 t h core cycles). Several activities have been performed in the reactor, to investigate the fuel integrity, among of them are: .fuel visual test with under water camera, which the results were recorder in the video cassette, primary water quality test during, reactor operation, fuel failure detector system examination and compared the PIE results in the Radiometallurgy Installation (RMI). The results showed that the fuel integrity, before and after irradiation, have still good performance and the fission products have not been released yet

Effect of post-digestion temperature on serial CSTR biogas reactor performance

DEFF Research Database (Denmark)

Boe, Kanokwan; Karakashev, Dimitar Borisov; Trably, Eric

2009-01-01

The effect of post-digestion temperature on a lab-scale serial continuous-flow stirred tank reactor (CSTR) system performance was investigated. The system consisted of a main reactor operated at 55 degrees C with hydraulic retention time (HRT) of 15 days followed by post-digestion reactors with HRT...

Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 15, 2011, 10:30 AM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du mardi 15 mars 2011 a 10h30

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-07-01

This situation note is established according to the information gained on March 15, 2011, at 10:30 AM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 of the Fukushima I site (Dai-ichi), 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.)

Light-water nuclear reactors

International Nuclear Information System (INIS)

Drevon, G.

1983-01-01

This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work [fr

High and low nightly running behavior associates with nucleus accumbens N-Methyl-d-aspartate receptor (NMDAR) NR1 subunit expression and NMDAR functional differences.

Science.gov (United States)

Grigsby, Kolter B; Kovarik, Cathleen M; Rottinghaus, George E; Booth, Frank W

2018-04-03

The extent to which N-Methyl-d-aspartate (NMDA) receptors facilitate the motivation to voluntarily wheel-run in rodents has yet to be determined. In so, we utilized female Wistar rats selectively bred to voluntarily run high (HVR) and low (LVR) nightly distances in order to examine if endogenous differences in nucleus accumbens (NAc) NMDA receptor expression and function underlies the propensity for high or low motivation to voluntarily wheel-run. 12-14 week old HVR and LVR females were used to examine: 1.) NAc mRNA and protein expression of NMDA subunits NR1, NR2A and NR2B; 2.) NMDA current responses in isolated medium spiny neurons (MSNs) and 3.) NMDA-evoked dopamine release in an ex vivo preparation of NAc punches. Expectedly, there was a large divergence in nightly running distance and time between HVR and LVR rats. We saw a significantly higher mRNA and protein expression of NR1 in HVR compared to LVR rats, while seeing no difference in the expression of NR2A or NR2B. There was a greater current response to a 500 ms application of 300 μM of NMDA in medium-spiny neurons isolated from the NAc HVR compared to LVR animals. On average, NMDA-evoked punches (50 μM of NMDA for 10 min) taken from HVR rats retained ∼54% of the dopamine content compared to their bilateral non-evoked sides, while evoked punches from LVR animals showed no statistical decrease in dopamine content compared to their non-evoked sides. Collectively, these data suggest a potential link between NAc NR1 subunit expression as well as NMDA function and the predisposition for nightly voluntary running behavior in rats. In light of the epidemic rise in physical inactivity, these findings have the potential to explain a neuro-molecular mechanism that regulates the motivation to be physically active. Copyright © 2018 Elsevier B.V. All rights reserved.

Sulfide toxicity kinetics of a uasb reactor

Directory of Open Access Journals (Sweden)

D. R. Paula Jr.

2009-12-01

Full Text Available The effect of sulfide toxicity on kinetic parameters of anaerobic organic matter removal in a UASB (up-flow anaerobic sludge blanket reactor is presented. Two lab-scale UASB reactors (10.5 L were operated continuously during 12 months. The reactors were fed with synthetic wastes prepared daily using glucose, ammonium acetate, methanol and nutrient solution. One of the reactors also received increasing concentrations of sodium sulfide. For both reactors, the flow rate of 16 L.d-1 was held constant throughout the experiment, corresponding to a hydraulic retention time of 15.6 hours. The classic model for non-competitive sulfide inhibition was applied to the experimental data for determining the overall kinetic parameter of specific substrate utilization (q and the sulfide inhibition coefficient (Ki. The application of the kinetic parameters determined allows prediction of methanogenesis inhibition and thus the adoption of operating parameters to minimize sulfide toxicity in UASB reactors.

Brief overview of American Nuclear Society's research reactor standards

International Nuclear Information System (INIS)

Richards, Wade J.

1984-01-01

The American Nuclear Society (ANS) established the research reactor standards group in 1968. The standards group, known as ANS-15, was established for the purpose of developing, preparing, and maintaining standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training

24. annual meeting of the International working group on fast reactors, Tsuruga, Japan, 15-18 April 1991

International Nuclear Information System (INIS)

1991-05-01

The agenda of the meeting included approval of the minutes of the 23rd meeting of the IWGFR, report of the Scientific Secretary regarding the activities of the Working Group, short presentation on national situation in nuclear energy and in the fast reactor field in particular, consideration of conferences on fast breeder reactors, consideration of arrangements for specialists' meetings in 1991, selection of topics for specialists' meetings to be held in 1992, consideration of joint research activities, informal discussion on ''The Economics of Fast Reactors''

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

RA research reactor - properties and experimental capabilities; Istrazivacki reaktor RA - Tehnicke karakteristike i eksploatacione mogucnosti

Energy Technology Data Exchange (ETDEWEB)

Milosevic, M; Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1978-05-15

The brief survey of the Reactor RA exploitation experience, as well as the reactor equipment state, after 18 years of operation is presented. The results of efforts spent on reactor characteristics improvement in order to ensure safe and reliable reactor operation for next 15-20 years, are described. Prikazani su fragmenti iz eksploatacije reaktora kao i stanje opreme, posle 18 godina rada. Na kraju je dat prikaz sta je preduzeto i sta se preduzima da se poboljsaju karakteristike i poveca sigurnost i bezbednost rada za sledecih 15-20 godina.

Operational and reliability experience with reactor instrumentation

International Nuclear Information System (INIS)

Dixon, F.; Gow, R.S.

1978-01-01

In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

Energy Technology Data Exchange (ETDEWEB)

Rubindamayugi, M S.T.; Salakana, L K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

1998-12-31

The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

Energy Technology Data Exchange (ETDEWEB)

Rubindamayugi, M.S.T.; Salakana, L.K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

1997-12-31

The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

Fusion reactor safety

International Nuclear Information System (INIS)

1987-12-01

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

Production of 15N-enriched nitric acid (H15NO3

Directory of Open Access Journals (Sweden)

C. R. Sant Ana Filho

2008-12-01

Full Text Available Techniques that employ 15N have proved to be an important tool in many areas of the agronomic and biomedical sciences. Nevertheless, their use is limited by methodological difficulties and by the price of compounds in the international market. Nitric compounds (15NO3- have attracted the interest of researchers. However, these compounds are not currently produced in Brazil. Thus, in the present work H15NO3 was obtained from the oxidation of anhydrous 15NH3. The method we used differs from the industrial process in that the absorption tower is replaced with a polytetrafluoroethylene-lined, stainless-steel hydration reactor. The process output was evaluated based on the following parameters: reaction temperature; ratio of reagents; pressure and flow of 15NH3(g through the catalyst (Pt/Rh. The results showed that, at the best conditions (500 ºC; 50 % excess O2; 0.4 MPa; and 3.39 g.min-1 of 15NH3, a conversion percentage (N-15NH3 to N-15NO3- of 62.2 %, an overall nitrogen balance (N-15NH3 + N-15NO3- of 86.8 %, and purity higher than 99 % could be obtained.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

Integrated Multiscale Latent Variable Regression and Application to Distillation Columns

Directory of Open Access Journals (Sweden)

Muddu Madakyaru

2013-01-01

Full Text Available Proper control of distillation columns requires estimating some key variables that are challenging to measure online (such as compositions, which are usually estimated using inferential models. Commonly used inferential models include latent variable regression (LVR techniques, such as principal component regression (PCR, partial least squares (PLS, and regularized canonical correlation analysis (RCCA. Unfortunately, measured practical data are usually contaminated with errors, which degrade the prediction abilities of inferential models. Therefore, noisy measurements need to be filtered to enhance the prediction accuracy of these models. Multiscale filtering has been shown to be a powerful feature extraction tool. In this work, the advantages of multiscale filtering are utilized to enhance the prediction accuracy of LVR models by developing an integrated multiscale LVR (IMSLVR modeling algorithm that integrates modeling and feature extraction. The idea behind the IMSLVR modeling algorithm is to filter the process data at different decomposition levels, model the filtered data from each level, and then select the LVR model that optimizes a model selection criterion. The performance of the developed IMSLVR algorithm is illustrated using three examples, one using synthetic data, one using simulated distillation column data, and one using experimental packed bed distillation column data. All examples clearly demonstrate the effectiveness of the IMSLVR algorithm over the conventional methods.

Upgrading of the research reactors FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1981-01-01

In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

Inter renewal travelling wave reactor with rotary fuel columns

International Nuclear Information System (INIS)

Terai, Yuzo

2016-01-01

To realize the COP21 decision, this paper proposes Inter Renewal Travelling Wave Reactor that bear high burn-up rate 50% and product TRU fuel efficiently. The reactor is based on 4S Fast Reactor and has Reactor Fuel Columns as fuel assemblies that equalize temperature in the fuel assembly so that fewer structure is need to restrain thermal transformation. To equalize burn-up rate of all fuel assemblies in the reactor, each rotary fuel column has each motor-lifter. The rotary fuel column has two types (Cylinder type and Heat Pipe type using natrium at 15 kPa which supply high temperature energy for Ultra Super Critical power plant). At 4 years cycle all rotary fuel columns of the reactor are renewed by the metallurgy method (vacuum re-smelting) and TRU fuel is gotten from the water fuel. (author)

High-Temperature Gas-Cooled Test Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Anaerobic horizontal flow reactor with polyethylene terephthalate as support material

Directory of Open Access Journals (Sweden)

Marcelo Muñoz

2016-06-01

Full Text Available A pilot anaerobic reactor was installed to remove the organic load of wastewater from dairy industry. It uses a bacterial inoculum previously acclimated to the substrate. It was disposed horizontally and filled with pieces of polyethylene terephthalate (PET, from plastic bottles. The reactor was operated at room temperature, during 100 days, in three phases: 1 the reactor was stabilized with volumetric organic load from 0.013 to 0.500 kg/day.m³; 2 the hydraulic retention time was of 1 day and the volumetric organic load of 3 kg/day.m³; 3 the volumetric organic load was incremented from 4 to 6.6 kg/day.m³ and the hydraulic retention time was 1 day. Organic material removal efficiencies was of 85%, and approximately 75% were obtained in the second and third phase, respectively. The Y value was 0.15, indicating that 0.15 kg of biomass were generated by kg of QDO supplied to the reactor. Finally, the biomass generated inside the reactor was analyzed, obtaining a value of 18868 mg/L, which is a higher value than those of conventional systems.

A complete fuel development facility utilizing a dual core TRIGA reactor system

Energy Technology Data Exchange (ETDEWEB)

Middleton, A; Law, G C [General Atomic Co., San Diego, CA (United States)

1974-07-01

A TRIGA Dual Core Reactor System has been chosen by the Romanian Government as the heart of a new fuel development facility which will be operated by the Romanian Institute for Nuclear Technologies. The Facility, which will be operational in 1976, is an integral part of the Romanian National Program for Power Reactor Development, with particular emphasis being placed on fuel development. The unique combination of a new 14 MW steady state TRIGA reactor, and the well-proven TRIGA Annular Core Pulsing Reactor (ACPR) in one below-ground reactor pool resulted in a substantial construction cost savings and gives the facility remarkable experimental flexibility. The inherent safety of the TRIGA fuel elements in both reactor cores means that a secondary containment building is not necessary, resulting in further construction cost savings. The 14 MW steady state reactor gives acceptably high neutron fluxes for long- term testing of various prototype fuel-cladding-coolant combinations; and the TRIGA ACPR high pulse capability allows transient testing of fuel specimens, which is so important for accurate prediction of the performance of power reactor fuel elements under postulated failure conditions. The 14 MW steady state reactor has one large and three small in-core irradiation loop positions, two large irradiation loop positions adjacent to the core face, and twenty small holes in the beryllium reflector for small capsule irradiation. The power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position approaching 3.0 x 10{sup 14} n/cm{sup 2}-sec. The ACPR has one large dry central experimental cavity which can be loaded at pool level through a shielded offset loading tube; a small diameter in-core flux trap; and an in-core pneumatically-operated capsule irradiation position. A peak pulse of 15,000 MW will yield a peak fast neutron flux in the central experimental cavity of about 1.5 x 10{sup 17} n/cm{sup 2}-sec. The pulse width at

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

Neoadjuvant administration of Semliki Forest virus expressing interleukin-12 combined with attenuated Salmonella eradicates breast cancer metastasis and achieves long-term survival in immunocompetent mice

International Nuclear Information System (INIS)

Kramer, M. Gabriela; Masner, Martín; Casales, Erkuden; Moreno, María; Smerdou, Cristian; Chabalgoity, José A.

2015-01-01

Metastatic breast cancer is a major cause of death among women worldwide; therefore efficient therapeutic strategies are extremely needed. In this work we have developed a gene therapy- and bacteria-based combined neoadjuvant approach and evaluated its antitumor effect in a clinically relevant animal model of metastatic breast cancer. 2×10 8 particles of a Semliki Forest virus vector expressing interleukin-12 (SFV-IL-12) and/or 2×10 7 units of an aroC − Samonella Typhimurium strain (LVR01) were injected into 4T1 tumor nodules orthotopically implanted in mice. Tumors were surgically resected and long-term survival was determined. IL-12 and interferon-γ were quantified by Enzyme-Linked ImmunoSorbent Assay, bacteria was visualized by inmunohistochemistry and the number of lung metastasis was calculated with a clonogenic assay. SFV-IL-12 and LVR01 timely inoculated and followed by surgical resection of tumors succeeded in complete inhibition of lethal lung metastasis and long-term survival in 90 % of treated mice. The combined therapy was markedly synergistic compared to each treatment alone, since SFV-IL-12 monotherapy showed a potent antiangiogenic effect, being able to inhibit tumor growth and extend survival, but could not prevent establishment of distant metastasis and death of tumor-excised animals. On the other hand, LVR01 alone also showed a significant, although limited, antitumor potential, despite its ability to invade breast cancer cells and induce granulocyte recruitment. The efficacy of the combined therapy depended on the order in which both factors were administered; inasmuch the therapeutic effect was only observed when SFV-IL-12 was administered previous to LVR01, whereas administration of LVR01 before SFV-IL-12 had negligible antitumor activity. Moreover, pre-treatment with LVR01 seemed to suppress SFV-IL-12 antiangiogenic effects associated to lower IL-12 expression in this group. Re-challenged mice were unable to reject a second 4T1 tumor

Small and medium power reactors 1987

International Nuclear Information System (INIS)

1987-12-01

This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

Small and medium power reactors 1987

Science.gov (United States)

1987-12-01

This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

Isothermal temperature reactivity coefficient measurement in TRIGA reactor

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Trkov, A.

2002-01-01

Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Jozef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the temperature range between 15 o C and 25 o C. All reactivity measurements were performed at almost zero reactor power to reduce or completely eliminate nuclear heating. Slow and steady temperature decrease was controlled using the reactor tank cooling system. In this way the temperatures of fuel, of moderator and of coolant were kept in equilibrium throughout the measurements. It was found out that TRIGA reactor core loaded with standard fuel elements with stainless steel cladding has small positive isothermal temperature reactivity coefficient in this temperature range.(author)

Neutron noise measurement technique in a coupled reactor

International Nuclear Information System (INIS)

Genoud, J.P.

1976-01-01

Describes work carried out on the swimming pool reactor at the Physikalisch-Technische Bundesanstalt at Braunschweig. The reactor has two multiplying zones, is light water moderated, with 90% enriched 235 U fuel. There is a D 2 0 reservoir between the two parts of the reactor. Signal/noise ratio obtained by means of ionisation chamber type neutron detectors of 10 -13 amp/u.f. sensitivity is of the order of 40 dB and band frequency 1.5 kHz. Spectral density of the interzone interaction energy was obtained by use of Fourier transforms, previously corrected by a Hanning window. (S.W.)

The future 700 MWe pressurized heavy water reactor

International Nuclear Information System (INIS)

Bhardwaj, S.A.

2006-01-01

The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

Directory of Open Access Journals (Sweden)

José Tavares de Sousa

2008-01-01

Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Proceedings of the Canadian Nuclear Society 15. annual conference

Energy Technology Data Exchange (ETDEWEB)

Huynh, H M [Hydro-Quebec, Montreal, PQ (Canada)

1994-12-31

The proceedings of the 15. annual conference of the Canadian Nuclear Society cover a wide range of nuclear topics, but the emphasis is on CANDU reactors and Canadian experience. The 89 papers are arranged in 17 sessions dealing with the following subjects: thermalhydraulics, fuel channels, operations, reactor physics, fuel, new technology, safety, training, waste management. The individual papers have been abstracted separately.

Proceedings of the Canadian Nuclear Society 15. annual conference

International Nuclear Information System (INIS)

Huynh, H.M.

1994-01-01

The proceedings of the 15. annual conference of the Canadian Nuclear Society cover a wide range of nuclear topics, but the emphasis is on CANDU reactors and Canadian experience. The 89 papers are arranged in 17 sessions dealing with the following subjects: thermalhydraulics, fuel channels, operations, reactor physics, fuel, new technology, safety, training, waste management. The individual papers have been abstracted separately

The Chernobyl reactor accident

International Nuclear Information System (INIS)

1986-01-01

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

The state of the PIK reactor construction

International Nuclear Information System (INIS)

Konoplev, K.A.

1995-01-01

Principle concepts of the PIK reactor project were stated late in the 60's but its construction was started in 1976. By the year 1986 the initial project was realised by approximately 70% but then, after Chernobyl accident the construction was essentially frozen to adjust the project to the revised nuclear safety regulations. The revised project was approved only in 1990 when the country was on the threshold of serious economic problems. The PIK reactor is a source of neutrons placed in the heavy water reflector. The fuel is uranium-235 (90% enrichment) of total weight 27 kg. Light water is used as moderator and coolant. Design parameters: thermal power is 100 W; thermal neutron flux in the reflector is 1.2x10 15 n/cm 2 s; in the central vertical beam tube is 5x10 15 n/cm 2 s; number of horizontal beam tubes is 10; diameter of beam tubes is 10 cm, with the possibility of replacement with beam tubes up to 25 cm in diameter. The reactor will be equipped with sources of hot, cold, and ultracold neutrons to obtain beams in different intervals of energy spectrum. The low temperature circuit will enable to irradiate samples at helium temperatures. The reactor has three series cooling circuits. Emergency core cooling systems in LOCA are double and in emergency power supply system is triple. The PIK reactor has no single common containment but four separate systems: for pipelines and units of the first circuit, for heavy water reflector, for operating hall, and for experimental beam tubes hall

Swelling and irradiation creep of neutron irradiated 316Ti and 15-15Ti steels

International Nuclear Information System (INIS)

Maillard, A.; Touron, H.; Seran, J.L.; Chalony, A.

1992-01-01

The global behavior, the swelling and irradiation creep resistances of cold worked 316Ti and 15-15Ti, two variants of austenitic steels in use as core component materials of the French fast reactors, are compared. The 15-15Ti leads to a significant improvement due to an increase in the incubation dose swelling. The same phenomena observed on 316Ti are found on 15-15Ti. All species without fuel like samples, wrappers or empty clad swell and creep less than fuel pin cladding irradiated in the same conditions. To explain the swelling difference, as for 316Ti, thermal gradient is also invoked but the irradiation creep difference is not yet clearly understood. To predict the behavior of clads it is indispensable to study the species themselves and to use specific rules. All results confirm the good behavior of 15-15Ti, the best behavior being obtained with the 1% Si doped version irradiated up to 115 dpa

ECP evaluation by water radiolysis and ECP model calculations

Energy Technology Data Exchange (ETDEWEB)

Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Rez (Czech Republic)

2010-07-01

In-pile ECP measurements data was evaluated by water radiolysis calculations. The data was obtained by using an in-pile loop in an experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Three types of ECP sensors, a Pt electrode, an Ag/AgCl sensor and a zirconia membrane sensor containing Ag/Ag{sub 2}O were used at several levels of the irradiation rig at various neutron flux and gamma rates. For water radiolysis calculation, the in-pile loop was modeled to several nodes following their design specifications, operating conditions such as flow rates, dose rate distributions of neutron and gamma-ray and so on. Concentration of chemical species along the water flow was calculated by a radiolysis code, WRAC-J. The radiolysis calculation results were transferred to an ECP model. In the model, anodic and cathodic current densities were calculated with combination of an electrochemistry model and an oxide film growth model. The measured ECP data were compared with the radiolysis/ECP calculation results, and applicability the of radiolysis model was confirmed. In addition, anomalous phenomenon appears in the in-pile loop was also investigated by radiolysis calculations. (author)

Use of sup(233)U for high flux reactors

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Liem, P.H.

1991-01-01

The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

Radiation streaming in power reactors. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Lahti, G.P.; Lee, R.R.; Courtney, J.C. (eds.)

1979-02-01

Separate abstracts are included for each of the 14 papers given at a special session on Radiation Streaming in Power Reactors held on November 15 at the American Nuclear Society 1978 Winter Meeting in Washington, D.C. The papers describe the methods of calculation, the engineering of shields, and the measurement of radiation environments within the containments of light water power reactors. Comparisons of measured and calculated data are used to determine the accuracy of computer predictions of the radiation environment. Specific computational and measurement techniques are described and evaluated. Emphasis is on radiation streaming in the annular region between the reactor vesel and the primary shield and its resultant environment within the primary containment.

Laparoscopic ventral rectopexy is effective for solitary rectal ulcer syndrome when associated with rectal prolapse.

Science.gov (United States)

Evans, C; Ong, E; Jones, O M; Cunningham, C; Lindsey, I

2014-03-01

Solitary rectal ulcer syndrome (SRUS) is uncommon and its management is controversial. The aim of this study was to evaluate the outcome of patients with SRUS who underwent laparoscopic ventral rectopexy (LVR). A review was performed of a prospective database at the Oxford Pelvic Floor Centre to identify patients between 2004 and 2012 with a histological diagnosis of SRUS. All were initially treated conservatively and surgical treatment was indicated only for patients with significant symptoms after failed conservative management. The primary end-point was healing of the ulcer. Secondary end-points included changes in the Wexner Constipation Score and Faecal Incontinence Severity Index (FISI). Thirty-six patients with SRUS were identified (31 women), with a median age of 44 (15–81) years. The commonest symptoms were rectal bleeding (75%) and obstructed defaecation (64%). The underlying anatomical diagnosis was internal rectal prolapse (n = 20), external rectal prolapse (n = 14) or anismus (n = 2). Twenty-nine patients underwent LVR and one a stapled transanal rectal resection (STARR) procedure. Nine (30%) required a further operation, six required posterior STARR for persistent SRUS and two a per-anal stricturoplasty for a narrowing at the healed SRUS site. Healing of the SRU was seen in 27 (90%) of the 30 patients and was associated with significant improvements in Wexner and FISI scores at a 3-year follow-up. Almost all cases of SRUS in the present series were associated with rectal prolapse. LVR resulted in successful healing of the SRUS with good function in almost all patients, but a significant number will require further surgery such as STARR for persistent obstructed defaecation.

PIK reactor construction status

International Nuclear Information System (INIS)

Konoplev, K.A.; Smolsky, S.L.

2001-01-01

The 100MW reactor PIK for fundamental researches has a thermal neutron flux of more than 10 15 n/cm 2 sec. This presentation outlines the construction state as of 2001, its prospects and completion tactics in the conditions of unstable finance. Construction of the reactor started in 1976. In 1986 construction of the building was completed and significant part of the installation work fulfilled. Construction of cooling systems was finished, the control panel assembled, and adjustment of the pump and gate valve control circuits started. After Chernobyl catastrophe, the USSR nuclear reactor safety requirements were revised. The PIK design did not meet these requirements and underwent considerable revision. The reconstruction design resulted in double the initial cost. Creation of the containment was the bulkiest part of the reconstruction. It brought about the need to disassemble the roofing of the building, dismantle all the equipment of the two upper floors, and lay up the equipment of the lower floors. As of 2001, construction in accordance with the revised design is at the stage of assemblage of the most important units, i.e. reactor itself, cooling system, heavy water system, and a number of auxiliary systems, such as depleted fuel storage, emergency cooling system etc. (orig.)

Core design of long life-cycle fast reactors operating without reactivity margin

International Nuclear Information System (INIS)

Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I.

2012-01-01

In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

Review of current literature related to GSI-15

International Nuclear Information System (INIS)

Lipinski, R.E.

1991-01-01

EG and G Idaho is evaluating the structural integrity of Reactor Pressure Vessel Supports as part of its resolution of NRC Generic Safety Issue (GSI) 15, 'Radiation effects on Reactor Vessel Supports'. Structural analyses have been performed by the NRC, ORNL, LLNL, Portland General Electric Co., and several consultants. Two methods of analyses were used. This paper contains a short discussion of the analyses and the merits of the two analysis methods. (author)

The advanced MAPLE reactor concept

International Nuclear Information System (INIS)

Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

1989-01-01

During the past several years, Atomic Energy of Canada Limited (AECL) has been developing the new MAPLE multipurpose reactor concept, which is capable of generating peak thermal neutron fluxes of up to 3 x 10 18 n/m 2 s in its heavy water reflector at a nominal thermal power level of 15MW. An assessment of the MAPLE-D 2 O reactor has shown that it could also be used as a high-flux neutron source. it could be developed to be used for several applications if a 12-site annular core is used. Thermal fluxes several times greater than in existing facilities would be available (author)

Final air test results for the 1/5-scale Mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Collins, E.K.; Lai, W.

1977-01-01

A loss-of-coolant accident (LOCA) in a boiling-water reactor (BWR) power plant has never occurred. However, because this type of accident is particularly severe, it is used as a principal basis for design. During a hypothetical LOCA in a Mark I BWR, air followed by steam is injected from a drywell into a toroidal wetwell about half-filled with water. A series of consistent, versatile, and accurate air-water tests simulating LOCA conditions was completed in the Lawrence Livermore Laboratory 1/5-Scale Mark I BWR Pressure Suppression Experimental Facility. Results from this test series were used to quantify the vertical loading function and to study the associated fluid dynamic phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variations of hydrodynamic-generated vertical loads with changes in drywell pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1/5-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings which are invariant. These groupongs show that if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor; the time when these forces occur is reduced by the square root of the scale factor

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Reactor physics activities in NEA member countries

International Nuclear Information System (INIS)

1990-01-01

This document is a compilation of National activity reports presented at the thirty-third Meeting of the NEA Committee on Reactor Physics, held at OECD Headquarters, Paris, from 15th - 19th October 1990

The need and prospects for improved fusion reactors

International Nuclear Information System (INIS)

Krakowski, R.A.; Hagenson, R.L.; Miller, R.L.

1986-01-01

Conceptual fusion reactor studies over the past 10-15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100-200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed

Fusion reactor design and technology 1986. V. 1

International Nuclear Information System (INIS)

1987-01-01

The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

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.

Thermionic nuclear reactor systems

International Nuclear Information System (INIS)

Kennel, E.B.

1986-01-01

Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

Training nuclear power plant personnel on SR-O reactor

International Nuclear Information System (INIS)

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

1983-01-01

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

Research reactors for the social safety and prosperous neutron use

International Nuclear Information System (INIS)

Ito, Yasuo

2000-01-01

The present status of nuclear reactors in Japan and the world was briefly described in this report. Aiming to construct a background of stable future society dependent on nuclear energy, the necessity to establish an organization for research reactors in Japan was pointed out. There are a total of 468 reactors in the world, but only 248 of them are running at present and most of them are superannuated. In Japan, 15 research reactors are running and 8 of them are under collaborative utilization, but not a few of them have various problems. In the education of atomic energy, a reactor is dispensable for understanding its working principle through practice learning. Furthermore, a research reactor has important roles for development of power reactor in addition to various basic studies such as activation analysis, fission track, biological irradiation, neutron scattering, etc. Application of a reactor has been also progressing in industrial and medical fields. However, operation of the reactors has become more and more difficult in Japan because of a large running cost and a lack of residential consensus for nuclear reactor. Here, the author proposed an establishment of organization of research reactor in order to promote utilization of a reactor in the field of education, rearing of professionals and science and engineering. (M.N.)

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

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

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Fast reactors - Dounreay and the future

International Nuclear Information System (INIS)

Jordan, G.

1988-01-01

In 1960 at Dounreay, the Dounreay Fast Reactor (DFR) supplied the world's first fast reactor grid electricity, and went on to a highly successful career as a test facility, as fuel designs advanced. In the 1960s, the Prototype Fast Reactor (PFR) was designed and built, beginning operation in 1974. The PFR was built to provide a sound technical and experienced base to support the UK's future Fast Reactor development and design. The in-vessel fuel handling facilities have demonstrated the flexibility of the pool design and a considerable body of in-core fuel handling experience is available. A key issue for further Fast Reactor application is the performance of fuel and, because PFR was designed to take full-scale fuel assemblies, the fuel performance experience is directly relevant to commercial designs. The original PFR design irradiation target of 60000 MWd/t U (equivalent to 7.5 % burn-up) has already been exceeded by a factor of more than two and a 15.9 % burn-up sub-assembly has been discharged and reprocessed without difficulty. Soon a 20 % sub-assembly will follow. Also the PFR reprocessing plant has demonstrated the safety and efficiency of this essential adjunct to Fast Reactor operation. The safety and the environmental protection features of both the PFR and its fuel reprocessing plant have been demonstrated over the last 14 years. 2 refs., 3 figs

Department of Energy's High Flux Beam Reactor (HFBR), September 15--19, 1980: An independent on-site safety review

International Nuclear Information System (INIS)

1981-02-01

The intent of this on-site safety review was to make a broad management assessment of HFBR operations, rather than conduct a detailed in-depth audit. The result of the review should only be considered as having identified trends or indications. The Team's observations and recommendations for the most part are based upon licensed reactor facility practices used to meet industry standards. These standards form the basis for many of the comments in this report. The Team believes that a uniform minimum standard of performance should be achieved in the operation of DOE reactors. In order to assure that this is accomplished, clear standards are necessary. Consistent with the past AEC and ERDA policy, the team has used the standards of the commercial nuclear power industry. It is recognized that this approach is conservative in that the HFBR reactor has a significantly greater degree of inherent safety (low pressure, temperature, power, etc.) than a licensed reactor

International symposium on research reactor utilization, safety and management. Book of extended synopses

International Nuclear Information System (INIS)

1999-01-01

The Symposium, considered as an important meeting of the owners and operators of research reactors as well as scientists concerned with problems of research reactors operation, management and safety covered the following topics: global and regional overview of research reactors, research reactors utilisation, research reactors safety, research reactors management, research reactors engineering. IAEA Research Reactors Database (RRDB) contains data concerning 291 operational research reactors, 247 shutdown reactors, 106 decommissioned reactors, 15 under construction and 15 new reactors planned. There is quite an even distribution of operational research reactors among 58 countries. Although about 66% of operational research reactors described in the RRDB are over 30 years old, the number of research reactors under construction or planned appears to have increased in recent years. According to the RRDB, the major applications of research reactors are in the field of neutron activation analysis, isotope production and neutron scattering work. Great concern was shown for several aspects of research reactors safety, especially since the average age of the operating research reactors is almost 30 years. Ageing problems involve more than the degradation of properties of the materials. Issues such as outdated equipment, lack of spare parts, outdating of the control and documentation systems related to the reactor, as well as budgetary limitations, affect the safety of some reactors. There are serious problems related to the spent fuel condition and the ageing of fuel storage facilities, in particular corrosion and leakage. The outstanding issues of concern are life extension of the spent fuel storage facilities and the future of take-back programmes of foreign research reactor fuels that will not be continued. A number of discussions related to safety requirements were focused on licensing and regulatory issues, especially in the case of older research reactors and those

International symposium on research reactor utilization, safety and management. Book of extended synopses

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-11-01

The Symposium, considered as an important meeting of the owners and operators of research reactors as well as scientists concerned with problems of research reactors operation, management and safety covered the following topics: global and regional overview of research reactors, research reactors utilisation, research reactors safety, research reactors management, research reactors engineering. IAEA Research Reactors Database (RRDB) contains data concerning 291 operational research reactors, 247 shutdown reactors, 106 decommissioned reactors, 15 under construction and 15 new reactors planned. There is quite an even distribution of operational research reactors among 58 countries. Although about 66% of operational research reactors described in the RRDB are over 30 years old, the number of research reactors under construction or planned appears to have increased in recent years. According to the RRDB, the major applications of research reactors are in the field of neutron activation analysis, isotope production and neutron scattering work. Great concern was shown for several aspects of research reactors safety, especially since the average age of the operating research reactors is almost 30 years. Ageing problems involve more than the degradation of properties of the materials. Issues such as outdated equipment, lack of spare parts, outdating of the control and documentation systems related to the reactor, as well as budgetary limitations, affect the safety of some reactors. There are serious problems related to the spent fuel condition and the ageing of fuel storage facilities, in particular corrosion and leakage. The outstanding issues of concern are life extension of the spent fuel storage facilities and the future of take-back programmes of foreign research reactor fuels that will not be continued. A number of discussions related to safety requirements were focused on licensing and regulatory issues, especially in the case of older research reactors and those

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Estudo piloto com Eco 3D das modificações geométricas do VE após infarto do miocárdio 3D Echo pilot study of geometric Left Ventricular changes after acute myocardial infarction

Directory of Open Access Journals (Sweden)

Marcelo Luiz Campos Vieira

2013-01-01

Full Text Available FUNDAMENTO: Remodelamento ventricular esquerdo (RVE após IAM caracteriza fator de mau prognóstico. Há pouca informação na literatura sobre o RVE analisado com ecocardiografia tridimensional (ECO 3D OBJETIVO: Analisar com ECO 3D as modificações geométricas e volumétricas do ventrículo esquerdo (VE seis meses após IAM em pacientes submetidos a tratamento primário percutâneo. MÉTODOS: Estudo prospectivo com ECO 3D de 21 indivíduos (16 homens, 56 ± 12 anos, acometidos por IAM com elevação do segmento ST. Foi feita a análise morfofuncional (VE com ECO 3D (volumes, FEVE, índice de esfericidade 3D até sete dias e seis meses após o IAM. RVE foi considerado para aumento > 15% do volume diastólico final do VE (VDFVE após seis meses do IAM, comparado ao VDFVE até sete dias do evento. RESULTADOS: Oito (38% pacientes apresentaram RVE. Medidas ecocardiográficas (n = 21 pacientes: I- até sete dias do IAM: 1- VDFVE: 92,3 ± 22,3 mL; 2- FEVE: 0,51 ± 0,01; 3- índice de esfericidade: 0,38 ± 0,05; II- após seis meses: 1- VDFVE: 107,3 ± 26,8 mL; 2- FEVE: 0,59 ± 0,01; 3- índice de esfericidade: 0,31 ± 0,05. Coeficiente de correlação (r entre índice de esfericidade até sete dias do IAM e VDFVE aos seis meses (n = 8 após o IAM: r: 0,74, p = 0,0007; (r entre índice de esfericidade após seis meses do IAM e VDFVE aos seis meses do IAM: r: 0,85, p BACKGROUND: Left ventricular remodeling (LVR after AMI characterizes a factor of poor prognosis. There is little information in the literature on the LVR analyzed with three-dimensional echocardiography (3D ECHO. OBJECTIVE: To analyze, with 3D ECHO, the geometric and volumetric modifications of the left ventricle (VE six months after AMI in patients subjected to percutaneous primary treatment. METHODS: Prospective study with 3D ECHO of 21 subjects (16 men, 56 ± 12 years-old, affected by AMI with ST segment elevation. The morphological and functional analysis (LV with 3D ECHO (volumes

TMI-2 reactor vessel plenum final lift

International Nuclear Information System (INIS)

Wilson, D.C.

1986-01-01

Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

The experimental program of neutronphysics for advanced water reactors

International Nuclear Information System (INIS)

Martin-Deider, L.; Cathalu, S.; Santamarina, A.; Gomit, M.

1985-11-01

The C.E.A. and E.D.F. has jointly undertaken a program of experimental studies on under-moderated water lattices, with mixed oxide fuel UO 2 -PuO 2 . Undermoderated lattices offer high conversion ratios. This type of lattice could limit in the future the natural uranium consumption of pressurized water reactors. This experimental program is aimed at qualifying neutron transport calculations in a large range of moderating ratio (between 0.5 and 1.5). It includes three experiments: ERASME, a critical experiment of large size in the EOLE reactor at Cadarache; ICARE, an irradiation experiment in the MELUSINE reactor at Grenoble; and an experiment to measure the reactivity effects by oscillations in the MINERVE reactor at Cadarache [fr

Predicting Structure-Function Relations and Survival following Surgical and Bronchoscopic Lung Volume Reduction Treatment of Emphysema.

Science.gov (United States)

Mondoñedo, Jarred R; Suki, Béla

2017-02-01

Lung volume reduction surgery (LVRS) and bronchoscopic lung volume reduction (bLVR) are palliative treatments aimed at reducing hyperinflation in advanced emphysema. Previous work has evaluated functional improvements and survival advantage for these techniques, although their effects on the micromechanical environment in the lung have yet to be determined. Here, we introduce a computational model to simulate a force-based destruction of elastic networks representing emphysema progression, which we use to track the response to lung volume reduction via LVRS and bLVR. We find that (1) LVRS efficacy can be predicted based on pre-surgical network structure; (2) macroscopic functional improvements following bLVR are related to microscopic changes in mechanical force heterogeneity; and (3) both techniques improve aspects of survival and quality of life influenced by lung compliance, albeit while accelerating disease progression. Our model predictions yield unique insights into the microscopic origins underlying emphysema progression before and after lung volume reduction.

Fast breeder reactor research

International Nuclear Information System (INIS)

1975-01-01

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

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

International Nuclear Information System (INIS)

Hien, P.D.

1999-01-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

Energy Technology Data Exchange (ETDEWEB)

Hien, P.D. [Vietnam Atomic Energy Agency, Hanoi (Viet Nam)

1999-08-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Space Fission Reactor Structural Materials: Choices Past, Present and Future

International Nuclear Information System (INIS)

Busby, Jeremy T.; Leonard, Keith J.

2007-01-01

Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

measurements of the absorption resonance integrals by reactor oscillator method

International Nuclear Information System (INIS)

Markovic, V.; Kocic, A.

1965-12-01

Experimental values of resonance integrals for silver vary significantly dependent on authors. That is why we have chosen this sample to measure RI. On the other hand, nuclear fuel (for example natural uranium) still represents an interesting objective for research in reactor physics. Measurements of natural uranium are done as a function of S/M. Measurements were done by amplitude reactor oscillator ROB-1/5 with precision from 0.5% - 2% dependent on the conditions of the oscillator. Measurements were completed at the heavy water reactor RB with 2% enriched uranium fuel [fr

Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

International Nuclear Information System (INIS)

Abou Yehia, H.; Bars, G.; Tran Dai

2001-01-01

After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

The reactor accident in the Three Mile Island-2 reactor plant. (Harrisburg, USA)

International Nuclear Information System (INIS)

Anon.

1979-01-01

The presentation of the accident development is based on the bulletin of the Atomic Industrial Forum (AIF) from the 6th of April, 1979. In addition, there are some short pieces of information, No. 14 and 15 of the association for reactor security as well as written and verbal information of the firms Brown, Boveri and Co/Babcock-Brown Boveri Reaktor GmbH (BBC/BBR). (orig.) [de

Anaerobic sewage treatment in a one-stage UASB reactor and a combined UASB Digester system

NARCIS (Netherlands)

Mahmoud, N.A.; Zeeman, G.; Gijzen, H.J.; Lettinga, G.

2004-01-01

The treatment of sewage at 15°C was investigated in a one-stage upflow anaerobic sludge blanket (UASB) reactor and a UASB-Digester system. The latter consists of a UASB reactor complemented with a digester for mutual sewage treatment and sludge stabilisation. The UASB reactor was operated at a

Synthesis of 15N-enriched urea (CO(15NH22 from 15NH3, CO, and S in a discontinuous process

Directory of Open Access Journals (Sweden)

C. R. Sant Ana Filho

2012-12-01

Full Text Available CO(15NH22 enriched with the stable isotope 15N was synthesized based on a reaction involving CO, 15NH3, and S in the presence of CH3OH. The method differs from the industrial method; a stainless steel reactor internally lined with polytetrafluoroethylene (PTFE was used in a discontinuous process under low pressure and temperature. The yield of the synthesis was evaluated as a function of the parameters: the amount of reagents, reaction time, addition of H2S, liquid solution and reaction temperature. The results showed that under optimum conditions (1.36, 4.01, and 4.48 g of 15NH3, CO, and S, respectively, 40 ml CH3OH, 40 mg H2S, 100 ºC and 120 min of reaction 1.82 g (yield 76.5% of the compound was obtained per batch. The synthesized CO(15NH22 contained 46.2% N, 0.55% biuret, melting point of 132.55 ºC and did not exhibit isotopic fractionation. The production cost of CO(15NH22 with 90.0 at. % 15N was US$ 238.60 per gram.

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Status of the advanced boiling water reactor and simplified boiling water reactor

International Nuclear Information System (INIS)

Smith, P.F.

1992-01-01

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

Microprocessor tester for the treat upgrade reactor trip system

International Nuclear Information System (INIS)

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations

Cavity temperature and flow characteristics in a gas-core test reactor

Science.gov (United States)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Socioeconomic consequences of nuclear reactor accidents

International Nuclear Information System (INIS)

Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

1984-06-01

This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables

Requirements for light water reactors

International Nuclear Information System (INIS)

Hedin, F.

2009-01-01

The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10 -6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

Nuclear Reactor RA Safety Report, Format and Contents

International Nuclear Information System (INIS)

1986-11-01

This is a new complete version of the safety report of nuclear reactor RA is made according to the recommendations of the IAEA. Report includes all the relevant data needed for evaluation of safe operation of this nuclear facility. Each of seven volumes of this report cover separate topics as follows: (1) introduction; (2) Site characteristics; (3) description of the reactor building and installations; (4) description of the reactor; (5) description of the coolant system; (6) description of the regulation and safety instrumentation; (7) description of the power supply system; (8) description of the auxiliary systems; (9) radiation protection issues; (10) radioactive waste management (11) reactor operation; (12) accident analysis during previous operation; (13) analysis of possible accident causes; (14) safety analysis and preventive actions: (15) analysis of significant accidents; (16) analysis of maximum possible accident; (17) environmental impact analysis in case of accident [sr

Benchmark physics tests in the metallic-fueled assembly ZPPR-15

International Nuclear Information System (INIS)

McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.

1989-01-01

Results of the first benchmark physics tests of a metallic-fueled, demonstration-size liquid-metal reactor (LMR) are reported. A simple, two-zone, cylindrical conventional assembly was built with three distinctly different compositions to represent the stages of the Integral Fast Reactor fuel cycle. Experiments included criticality, control, power distribution, reaction rate ratios, reactivity coefficients, shielding, kinetics, and spectrum. Analysis was done with three-dimensional nodal diffusion calculations and ENDF/B-V.2 cross sections. Predictions of the ZPPR-15 reactor physics parameters agreed sufficiently well with the measured values to justify confidence in design analyses for metallic-fueled LMRs

Benchmark physics tests in the metallic-fuelled assembly ZPPR-15

International Nuclear Information System (INIS)

McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.

1987-01-01

Results of the first benchmark physics tests of a metallic-fueled, demonstration-size, liquid metal reactor are reported. A simple, two-zone, cylindrical conventional assembly was built with three distinctly different compositions to represent the stages of the Integral Fast Reactor fuel cycle. Experiments included criticality, control, power distribution, reaction rate ratios, reactivity coefficients, shielding, kinetics and spectrum. Analysis was done with 3-D nodal diffusion calculations and ENDFIB-V.2 cross sections. Predictions of the ZPPR-15 reactor physics parameters agreed sufficiently well with the measured values to justify confidence in design analyses for metallic-fueled LMRs

Steady-state plasma and reactor parameters for elliptical cross section tokamaks with very large power ratings

International Nuclear Information System (INIS)

Usher, J.L.; Powell, J.R.

1975-06-01

In previous studies only circular cross section reactor plasmas were considered. The purpose of this research is to examine the effects of elliptical plasma cross sections. Several technological benefits have been determined. Maximum magnetic field strength requirements are 30 to 65 percent less than for 5000 MW (th) reactors and may be as much as 40 percent less than for circular cross section reactors of comparable size. Very large n tau values are found (10 15 to 10 17 sec/cm 3 ), which produce large burn-up fractions (15 to 60 percent). There is relatively little problem with impurity build-up. Long confinement times (60 to 500 seconds) are found. Finally, the elliptical cross section reactors exhibit a major toroidal radius reduction of as large as 30 percent over circular reactors operating at comparable power levels

Lung volume reduction in pulmonary emphysema from the radiologist's perspective; Lungenvolumenreduktion beim Lungenemphysem aus der Sicht des Radiologen

Energy Technology Data Exchange (ETDEWEB)

Doellinger, F.; Poellinger, A. [Charite Universitaetsmedizin Berlin (Germany). Dept. of Radiology; Huebner, R.H. [Charite Universitaetsmedizin Berlin (Germany). Dept. of Internal Medicine/Infectious and Respiratory Diseases; Kuhnigk, J.M. [Fraunhofer MEVIS, Bremen (Germany). Inst. for Medical Image Computing

2015-08-15

Pulmonary emphysema causes decrease in lung function due to irreversible dilatation of intrapulmonary air spaces, which is linked to high morbidity and mortality. Lung volume reduction (LVR) is an invasive therapeutical option for pulmonary emphysema in order to improve ventilation mechanics. LVR can be carried out by lung resection surgery or different minimally invasive endoscopical procedures. All LVR-options require mandatory preinterventional evaluation to detect hyperinflated dysfunctional lung areas as target structures for treatment. Quantitative computed tomography can determine the volume percentage of emphysematous lung and its topographical distribution based on the lung's radiodensity. Modern techniques allow for lobebased quantification that facilitates treatment planning. Clinical tests still play the most important role in post-interventional therapy monitoring, but CT is crucial in the detection of postoperative complications and foreshadows the method's high potential in sophisticated experimental studies. Within the last ten years, LVR with endobronchial valves has become an extensively researched minimally-invasive treatment option. However, this therapy is considerably complicated by the frequent occurrence of functional interlobar shunts. The presence of ''collateral ventilation'' has to be ruled out prior to valve implantations, as the presence of these extraanatomical connections between different lobes may jeopardize the success of therapy. Recent experimental studies evaluated the automatic detection of incomplete lobar fissures from CT scans, because they are considered to be a predictor for the existence of shunts. To date, these methods are yet to show acceptable results.

Growth trends, genotype-environment interaction and genetic gains in six-year-old rubber tree clones (Hevea in São Paulo State, Brazil

Directory of Open Access Journals (Sweden)

Gonçalves Paulo de Souza

1998-01-01

Full Text Available Budwood from seven different clones of rubber tree [Hevea brasiliensis (Willd. ex Adr. de Juss. Müell. Arg.] was planted in replicated trials in four different test sites in the plateau region of the state of São Paulo, Brazil. The clones represented a range of imported germplasm, whereas the locations were selected to represent a range of rubber growing areas. Girths were measured for six years consecutively, before the initiation of tapping for latex. Total number of latex vessel rings (lvr and bark thickness (bt were measured at six years. The largest average for all characters was observed in Matão. Overall means for girth, total number of latex vessel rings and bark thickness at six years were 35.26 cm, 11.30 units and 4.83 mm, respectively. Both the test sites and clones showed statistically significant differences in girth, lvr and bt values. There were significant clone x site interactions. Girth at one year was not a reliable predictor of future field performance. Broad sense heritability for girth was 0.16, and for lvr and bt it was 0.28 and 0.40, respectively, at six years. Repeatability was quite high, from 0.52 to 0.75 for all characters in all years. When two clones out of seven were selected, expected genetic gain in girth was about 2.0%; for lvr and bt it was 7.0% and 14.6%, respectively.

Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

1990-01-01

Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

Conceptual design of D-3He FRC reactor 'ARTEMIS'

International Nuclear Information System (INIS)

Momota, H.; Ishida, A.; Kohzaki, Y.

1991-07-01

A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

The relationship between natural uranium and advanced fuel cycles in CANDU reactors

International Nuclear Information System (INIS)

Lane, A.D.; McDonnell, F.N.; Griffiths, J.

1988-11-01

CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

Compact-toroid fusion reactor based on the field-reversed theta pinch: reactor scaling and optimization for CTOR

International Nuclear Information System (INIS)

Hagenson, R.L.; Krakowski, R.A.

1980-01-01

Early scoping studies based on approximate, analytic models have been extended on the basis of a dynamic plasma model and an overall systems approach to examine a Compact Toroid (CT) reactor embodiment that uses a Field-Reversed Theta Pinch as a plasma source. The field-reversed plasmoid would be formed and compressionally heated to ignition prior to injection into and translation through a linear burn chamber, thereby removing the high-technology plasmoid source from the hostile reactor environment. Stabilization of the field-reversed plasmoid would be provided by a passive conduction shell located outside the high-temperature blanket but within the low-field superconducting magnets and associated radiation shielding. On the basis of this batch-burn but thermally steady-state approach a reactor concept emerges with a length below approx. 40 m that generates 300 to 400 MWe of net electrical power with a recirculating power fraction less than 0.15

Refurbishing the BR2 materials testing reactor

International Nuclear Information System (INIS)

Baugnet, J.M.; Dekeyser, J.; Gubel, P.

1995-01-01

SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

Simultaneous hydrogen utilization and in situ biogas upgrading in an anaerobic reactor

DEFF Research Database (Denmark)

Luo, Gang; Johansson, Sara; Boe, Kanokwan

2012-01-01

. The methane production rate of the reactor with H2 addition was 22% higher, compared to the control reactor only fed with manure. The CO2 content in the produced biogas was only 15%, while it was 38% in the control reactor. However, the addition of hydrogen resulted in increase of pH (from 8.0 to 8.3) due......The possibility of converting hydrogen to methane and simultaneous upgrading of biogas was investigated in both batch tests and fully mixed biogas reactor, simultaneously fed with manure and hydrogen. Batch experiments showed that hydrogen could be converted to methane by hydrogenotrophic...

Studies of tokamak fusion reactor dynamics. Progress report, June 1, 1975--February 15, 1976

International Nuclear Information System (INIS)

Mills, R.G.; Gralnick, S.L.

1976-01-01

An investigation of the effect of plasma shape and position on the inductive coupling between the plasma and the external poloidal field coils is briefly described. Research on a multi-node time-dependent point kinetics code with which to study the operating dynamics of a tokamak reactor is also mentioned

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Accident at the zero power reactor which happened on October 15 1958; Sur l'accident avec le reacteur de puissance zero du 15 octobre 1958

Energy Technology Data Exchange (ETDEWEB)

Savic, P [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

1959-03-15

During an experiment on the zero power heavy water reactor with natural uranium fuel in the Boris Kidric Institute of Nuclear Sciences, the reactor escaped control. Six staff members in the immediate surrounding of the bare assembly were exposed to high neutron and ionising irradiation. Other two employees who were at some bigger distance were exposed to doses higher than permitted. This paper deals with the circumstances that caused the accident, status of the dosimetry, control and alarm systems. Individual exposure doses were estimated according to the calculated neutron flux values obtained from measuring the activities of personal belongings made of gold and copper as well as radioactive phosphorous from urine.

Using cathode spacers to minimize reactor size in air cathode microbial fuel cells

KAUST Repository

Yang, Qiao

2012-04-01

Scaling up microbial fuel cells (MFCs) will require more compact reactor designs. Spacers can be used to minimize the reactor size without adversely affecting performance. A single 1.5mm expanded plastic spacer (S1.5) produced a maximum power density (973±26mWm -2) that was similar to that of an MFC with the cathode exposed directly to air (no spacer). However, a very thin spacer (1.3mm) reduced power by 33%. Completely covering the air cathode with a solid plate did not eliminate power generation, indicating oxygen leakage into the reactor. The S1.5 spacer slightly increased columbic efficiencies (from 20% to 24%) as a result of reduced oxygen transfer into the system. Based on operating conditions (1000ς, CE=20%), it was estimated that 0.9Lh -1 of air would be needed for 1m 2 of cathode area suggesting active air flow may be needed for larger scale MFCs. © 2012 Elsevier Ltd.

Effects of the reduction of the hydraulic retention time to 1.5 days at constant organic loading in CSTR, ASBR, and fixed-bed reactors – Performance and methanogenic community composition

International Nuclear Information System (INIS)

Schmidt, Thomas; Ziganshin, Ayrat M.; Nikolausz, Marcell; Scholwin, Frank; Nelles, Michael; Kleinsteuber, Sabine; Pröter, Jürgen

2014-01-01

The hydraulic retention time (HRT) is one of the key parameters in biogas processes and often it is postulated that a minimum HRT of 10–25 days is obligatory in continuous stirred tank reactors (CSTR) to prevent a washout of slow growing methanogens. In this study the effects of the reduction of the HRT from 6 to 1.5 days on performance and methanogenic community composition in different systems with and without immobilization operated with simulated thin stillage (STS) at mesophilic conditions and constant organic loading rates (OLR) of 10 g L −1 d −1 of volatile solids were investigated. With the reduction of the HRT process instability was first observed in the anaerobic sequencing batch reactor (ASBR) (at HRT of 3 days) followed by the CSTR (at HRT of 2 days). The fixed bed reactor (FBR) was stable until the end of the experiment, but the reduction of the HRT to 1.5 days caused a decrease of the specific biogas production to about 450 L kg −1 of VS compared to about 600 L kg −1 of VS at HRTs of 4–5 days. Methanoculleus and Methanosarcina were the dominant genera under stable process conditions in the CSTR and the ASBR and members of Methanosaeta and Methanospirillum were only present at HRT of 4 days and lower. In the effluent of the FBR Methanosarcina spp. were not detected and Methanosaeta spp. were more abundant then in the other reactors. - Highlights: • A CSTR was operated at high OLR of 10 (g L −1  d −1  VS) and low HRT of 3 days. • Exceeding washout of methanogenic archaea did not take place. • pH and nutrient concentrations influenced the reproduction rate more than HRT. • Methanoculleus and Methanosarcina were the dominant genera in the CSTR

Decommissioning of fast reactors after sodium draining

International Nuclear Information System (INIS)

2009-11-01

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

Reaction kinetic analysis of reactor surveillance data

Energy Technology Data Exchange (ETDEWEB)

Yoshiie, T., E-mail: yoshiie@rri.kyoto-u.ac.jp [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka-fu 590-0494 (Japan); Kinomura, A. [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka-fu 590-0494 (Japan); Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan)

2017-02-15

In the reactor pressure vessel surveillance data of a European-type pressurized water reactor (low-Cu steel), it was found that the concentration of matrix defects was very high, and a large number of precipitates existed. In this study, defect structure evolution obtained from surveillance data was simulated by reaction kinetic analysis using 15 rate equations. The saturation of precipitation and the growth of loops were simulated, but it was not possible to explain the increase in DBTT on the basis of the defect structures. The sub-grain boundary segregation of solutes was discussed for the origin of the DBTT increase.

Radiation protection personnel training in Research Reactors

International Nuclear Information System (INIS)

Fernandez, Carlos Dario; Lorenzo, Nestor Pedro de

1996-01-01

The RA-6 research reactor is considering the main laboratory in the training of different groups related with radiological protection. The methodology applied to several courses over 15 years of experience is shown in this work. The reactor is also involved in the construction, design, start-up and sell of different installation outside Argentina for this reason several theoretical and practical courses had been developed. The acquired experience obtained is shown in this paper and the main purpose is to show the requirements to be taken into account for every group (subjects, goals, on-job training, etc) (author)

RA reactor exploitation, task 3.08/01; Zadatak 3.08/01 - Eksploatacija reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report.

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

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

Microbiological airway colonization in COPD patients with severe emphysema undergoing endoscopic lung volume reduction

Directory of Open Access Journals (Sweden)

Trudzinski FC

2017-12-01

Full Text Available Franziska C Trudzinski,1 Frederik Seiler,1 Heinrike Wilkens,1 Carlos Metz,1 Annegret Kamp,1 Robert Bals,1 Barbara Gärtner,2 Philipp M Lepper,1 Sören L Becker2–4 1Department of Internal Medicine V – Pneumology, Allergology and Critical Care Medicine, ECLS Center Saar, University Medical Center Saarland and Saarland University, 2Institute of Medical Microbiology and Hygiene, Saarland University, Homburg/Saar, Germany; 3Swiss Tropical and Public Health Institute, 4University of Basel, Basel, Switzerland Background: Endoscopic lung volume reduction (eLVR is a therapeutic option for selected patients with COPD and severe emphysema. Infectious exacerbations are serious events in these vulnerable patients; hence, prophylactic antibiotics are often prescribed postinterventionally. However, data on the microbiological airway colonization at the time of eLVR are scarce, and there are no evidence-based recommendations regarding a rational antibiotic regimen.Objective: The aim of this study was to perform a clinical and microbiological analysis of COPD patients with advanced emphysema undergoing eLVR with endobronchial valves at a single German University hospital, 2012–2017.Patients and methods: Bronchial aspirates were obtained prior to eLVR and sent for microbiological analysis. Antimicrobial susceptibility testing of bacterial isolates was performed, and pathogen colonization was retrospectively compared with clinical parameters.Results: At least one potential pathogen was found in 47% (30/64 of patients. Overall, Gram-negative bacteria constituted the most frequently detected pathogens. The single most prevalent species were Haemophilus influenzae (9%, Streptococcus pneumoniae (6%, and Staphylococcus aureus (6%. No multidrug resistance was observed, and Pseudomonas aeruginosa occurred in <5% of samples. Patients without microbiological airway colonization showed more severe airflow limitation, hyperinflation, and chronic hypercapnia compared

Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

Energy Technology Data Exchange (ETDEWEB)

Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-08-28

This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

Brayton rotating units for space reactor power systems

Energy Technology Data Exchange (ETDEWEB)

Gallo, Bruno M.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies and Chemical and Nuclear Engineering Dept., The Univ. of New Mexico, Albuquerque, NM 87131 (United States)

2009-09-15

Designs and analyses models of centrifugal-flow compressor and radial-inflow turbine of 40.8kW{sub e} Brayton Rotating Units (BRUs) are developed for 15 and 40 g/mole He-Xe working fluids. Also presented are the performance results of a space power system with segmented, gas cooled fission reactor heat source and three Closed Brayton Cycle loops, each with a separate BRU. The calculated performance parameters of the BRUs and the reactor power system are for shaft rotational speed of 30-55 krpm, reactor thermal power of 120-471kW{sub th}, and turbine inlet temperature of 900-1149 K. With 40 g/mole He-Xe, a power system peak thermal efficiency of 26% is achieved at rotation speed of 45 krpm, compressor and turbine inlet temperatures of 400 and 1149 K and 0.93 MPa at exit of the compressor. The corresponding system electric power is 122.4kW{sub e}, working fluid flow rate is 1.85 kg/s and the pressure ratio and polytropic efficiency are 1.5% and 86.3% for the compressor and 1.42% and 94.1% for the turbine. For the same nominal electrical power of 122.4kW{sub e}, decreasing the molecular weight of the working fluid (15 g/mole) decreases its flow rate to 1.03 kg/s and increases the system pressure to 1.2 MPa. (author)

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Fission product chemistry in severe nuclear reactor accidents

International Nuclear Information System (INIS)

Nichols, A.L.

1990-09-01

A specialist's meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions)

Studies on fuel failure detection in Rikkyo Research Reactor

International Nuclear Information System (INIS)

Matsuura, T.; Hayashi, S.H.; Harasawa, S.; Tomura, K.

1992-01-01

Studies on fuel failure detection have been made since 1986 in Rikkyo Research Reactor. One of the methods is the monitoring of the trace concentration of fission products appearing in the air on the surface of the water tank of the reactor. The interested radionuclides here are 89 Rb and 138 Cs, which are the daughter nuclides of the FP rare gas nuclides, 89 Kr and 138 Xe, respectively and have the half lives of 15.2 min and 32.2 min respectively. They are detected on a filter paper attached on a conventional dust sampler, by sucking the air of the surface of the water for 15 ∼ 30 min during reactor operation (100 kW). In this presentation are reported the results of an attempt to increase the sensitivity of detecting these nuclides by introducing nitrogen gas bubbles into the water. The bubbling of the gas increased the sensitivity as much as several times compared with the case without bubbling. These measurements are giving us the 'background' concentration, the order of which is almost unchanged for these several years, --in 10 -6 Bq/cm 3 . The origin of these nuclides is considered to be not from the fuel but from the uranium contained as an impurity in the reactor material in the core. (author)

On blanket concepts of the Helias reactor

International Nuclear Information System (INIS)

Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

1999-07-01

The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

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

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Siloe, Osiris, and the future perspective of swimming-pool reactors

International Nuclear Information System (INIS)

Chatoux, J.; Denielou, G.; Lerouge, B.

1964-01-01

Siloe and Osiris are two new general purpose research reactors of the 'Commissariat a l'energie Atomique'. Siloe, located within the 'Centre d'Etudes Nucleaires' of Grenoble is a swimming pool reactor of the same type as Melusine and Triton. It operates, at a nominal power of 15 MW thermal and has reached the peak power of 20 MW thermal with two thirds of its cooling system working. The fast flux above 1 MeV, which is maximum at the center of the core at 15 MW thermal is 1,2. 10 14 . The core, quite open, is downward cooled. Average specific power is 159 kW/l. Osiris is under construction at Saclay. Designed for 50 MW thermal, this reactor is upward cooled. The fast flux at the center of the core above 1 MeV is calculated to be 2, 5.10 14 . The average designed specific power is 280 kW/l. A fixed zircaloy gamma shield makes a box round the core. Future perspectives open to non-pressurised swimming-pool reactors are examined. Ways are suggested for neutronic; thermal and shielding modifications which make possible further improvements in the performances and economy of these devices. (authors) [fr

CFD simulation on reactor flow mixing phenomena

International Nuclear Information System (INIS)

Kwon, T.S.; Kim, K.H.

2016-01-01

A pre-test calculation for multi-dimensional flow mixing in a reactor core and downcomer has been studied using a CFD code. To study the effects of Reactor Coolant Pump (RCP) and core zone on the boron mixing behaviors in a lower downcomer and core inlet, a 1/5-scale CFD model of flow mixing test facility for the APR+ reference plant was simulated. The flow paths of the 1/5-scale model were scaled down by the linear scaling method. The aspect ratio (L/D) of all flow paths was preserved to 1. To preserve a dynamic similarity, the ratio of Euler number was also preserved to 1. A single phase water flow at low pressure and temperature conditions was considered in this calculation. The calculation shows that the asymmetric effect driven by RCPs shifted the high velocity field to the failed pump's flow zone. The borated water flow zone at the core inlet was also shifted to the failed RCP side. (author)

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

State of development of high temperature gas-cooled reactors in foreign countries

International Nuclear Information System (INIS)

Sudo, Yukio

1990-01-01

Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

Small reactors with simplified design. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-11-01

There is a potential future need for small reactors for applications such as district heating, electricity production at remote locations and desalination. Nuclear energy can provide an environmentally benign alternative to meet these needs. For successful deployment, small reactors must satisfy the requirements of users, regulators and the general public. The IAEA has been following the developments in the field of small reactors as a part of the sub-programme on advanced reactor technology. In accordance with the interests of Member States, a Technical Committee meeting (TCM) was organized in Mississauga, Ontario, Canada, 15-19 May 1995 to discuss the status of designs and design requirements related to small reactors for diverse applications. The papers presented at the TCM and a summary of the discussions are contained in this TECDOC which, it is hoped, will serve the Member States as a useful source of technical information on the development of small reactors with simplified design. Refs, figs, tabs.

Small reactors with simplified design. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1997-11-01

There is a potential future need for small reactors for applications such as district heating, electricity production at remote locations and desalination. Nuclear energy can provide an environmentally benign alternative to meet these needs. For successful deployment, small reactors must satisfy the requirements of users, regulators and the general public. The IAEA has been following the developments in the field of small reactors as a part of the sub-programme on advanced reactor technology. In accordance with the interests of Member States, a Technical Committee meeting (TCM) was organized in Mississauga, Ontario, Canada, 15-19 May 1995 to discuss the status of designs and design requirements related to small reactors for diverse applications. The papers presented at the TCM and a summary of the discussions are contained in this TECDOC which, it is hoped, will serve the Member States as a useful source of technical information on the development of small reactors with simplified design

The safety of the fast reactor

International Nuclear Information System (INIS)

Matthews, R.R.

1977-01-01

Verbatim of an address by R.R. Matthews, Chief Nuclear Health and Safety Officer, UK Central Electricity Generating Board given on January 15th 1977. The object of this address was to give some opinions on the safety issues of fast reactors as seen from an operational point of view. An outline of the basic responsibilities for nuclear safety is first given, and it is emphasized that the Central Electricity Generating Board has a statutory responsibility for the safe operation of its nuclear plant. The Nuclear Installations Act places absolute responsibility on the operator for ensuring that injury to persons and damage to property do not occur, and the new Health and Safety at Work Act does likewise. In addition the Board has a Nuclear Health and Safety Department that has to ensure that adequate provision for safety is made in the design, construction, and operation of nuclear plant, and safety at operational stations is monitored continuously by inspectors. In addition the requirements of the Nuclear Installations Inspectorate, laid down in the site licence conditions, must be satisfied. All these requirements are here discussed in the light of application to commercial fast reactors. It is considered that the hazards to fast reactor operating personnel are small and little different from those of other types of reactor, and in some respects the fast reactor has advantages, particularly in regard to the use of a Na coolant. The possibility of various types of accident is considered. Radioactive effluent discharge is also considered. The fast reactor as an international problem is discussed, including security matters. The extensive experience gained in operation of the experimental and prototype fast reactors at Dounreay is emphasized. (U.K.)

TRIGA 14 MW Research Reactor Status and Utilization

International Nuclear Information System (INIS)

Barbos, D.; Ciocanescu, M.; Paunoiu, C.

2016-01-01

Institute for Nuclear Research is the owner of the largest family TRIGA research reactor, TRIGA14 MW research reactor. TRIGA14 MW reactor was designed to be operated with HEU nuclear fuel but now the reactor core was fully converted to LEU nuclear fuel. The full conversion of the core was a necessary step to ensure the continuous operation of the reactor. The core conversion took place gradually, using fuel manufactured in different batches by two qualified suppliers based on the same well qualified technology for TRIGA fuel, including some variability which might lead to a peculiar behaviour under specific conditions of reactor utilization. After the completion of the conversion a modernization program for the reactor systems was initiated in order to achieve two main objectives: safe operation of the reactor and reactor utilization in a competitive environment to satisfy the current and future demands and requirements. The 14 MW TRIGA research reactor operated by the Institute for Nuclear Research in Pitesti, Romania, is a relatively new reactor, commissioned 37 years ago. It is expected to operate for another 15-20 years, sustaining new fuel and testing of materials for future generations of power reactors, supporting radioisotopes production through the development of more efficient new technologies, sustaining research or enhanced safety, extended burn up and verification of new developments concerning nuclear power plants life extension, to sustain neutron application in physics research, thus becoming a centre for instruction and training in the near future. A main objective of the TRIGA14MW research reactor is the testing of nuclear fuel and nuclear material. The TRIGA 14 MW reactor is used for medical and industrial radioisotopes production ( 131 I, 125 I, 192 Ir etc.) and a method for 99 Mo- 99 Tc production from fission is under development. For nuclear materials properties investigation, neutron radiography methods have been developed in the INR. The

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

Science.gov (United States)

Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

2018-04-01

1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

The developments of anaerobic baffled reactor for wastewater ...

African Journals Online (AJOL)

STORAGESEVER

2010-03-15

Mar 15, 2010 ... as a series of up-flow anaerobic sludge blanket reactors. (UASBs). As the .... to determine whether the rate of gas production and viscosity affected .... behaviour of glucose fed PABR and used to examine the behaviour of the ...

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

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

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

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

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

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

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

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

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

Occupational radiation exposure at commercial nuclear power reactors, 1978

International Nuclear Information System (INIS)

Brooks, B.G.

1979-11-01

An updated compilation is presented of occupational radiation exposures at commercial nuclear power reactors for the years 1969 through 1978. Data received from the 64 light water cooled reactors (LWRs) that had completed at least one year of commercial operation as of December 31, 1978 are included. This represents an increase of seven reactors over the number contained in last year's report. The total number of personnel monitored at LWRs during 1978 increased by approximately 12% to 76,121. The number of workers that received measurable doses, however, increased only 8% to 45,978. The total collective dose for 1978 is estimated to be 31,806 man-rems, a small decrease from last year's value of 32,511, which results in the average dose per worker decreasing slightly to 0.69 rems. The average collective dose per reactor also decreased, by approximately 15%, to a value of 497 man-rems

9 CFR 147.15 - Laboratory procedure recommended for the bacteriological examination of mycoplasma reactors. 11

Science.gov (United States)

2010-01-01

... Laboratory procedure recommended for the bacteriological examination of mycoplasma reactors. 11 11 Yoder, H. W., Jr., “Mycoplasmosis.” In: Isolation and Identification of Avian Pathogens. (Stephen B. Hitchner, Chairman, Charles H. Domermuth, H. Graham Purchase, James E. Williams.) 1980, pp. 40-42, Creative Printing...

Measurement of the physics properties of gas-cooled fast reactors in the zero energy reactor PROTEUS and analysis of the results

International Nuclear Information System (INIS)

Richmond, R.

1982-12-01

The main aim of the fast reactor physics measurements carried out in the zero energy reactor PROTEUS was to check the performance of data sets and calculation methods used in the design of fast breeder reactors. This allowed the accuracy of the power reactor calculations to be determined and enabled an assessment to be made of whether this accuracy would be sufficient to allow the design, construction and licensing of the GCFR power reactor. In order to carry out the physics measurements an existing zero energy reactor was converted to a form in which a central fast reactor lattice was surrounded by thermal zones to drive the reactor critical. One of the most important measuring techniques used to check the performance of data sets and calculation methods was the determination of reaction rate ratios and, by using an appropriate range of nuclides, it was possible to obtain a detailed picture covering 70% of reactions taking place in the central part of the fast reactor zone and with an accuracy of +-1.5% in a typical ratio. A further technique used during the work on GCFR-PROTEUS was the measurement of neutron spectrum which was carried out in a wide range of environments and, in the later stages of the work, covered the energy range from 9 keV to 2.3 MeV. These measurements, in particular, indicated significant errors in the FGL4 scattering cross-sections. A third technique, which was developed to a high degree of accuracy, was the measurement of reactivity worths. This was used in measurements of the worths of small samples and also in the application of the null reactivity technique to determine k-infinity and hence the absorption cross-sections of reactor structural materials. (Auth.)

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

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

Ageing management practice in Fast Breeder Test Reactor

International Nuclear Information System (INIS)

Srinivasan, G.; Ramanathan, V.; Swaminathan, P.R.; Babu, A.; Rajasekarappa, E.; Rajendran, B.; Ramalingam, P.V.

2006-01-01

Fast Breeder Test Reactor is a 40 MWt, sodium cooled, PuC-UC fuelled fast reactor, located at Kalpakkam, India. The reactor went critical in October 85 with Mark I core rated for 10.5 MWt at a peak LHR of 320 W/cm. The reactor core was progressively enlarged and TG was synchronized to the grid in July 97. The present core has 41 fuel subassemblies rated for 15.7 MWt at a peak LHR of 320 W/cm. The reactor has so far been operated for 33000 h and has seen 660 EFPD of operation corresponding to peak LHR of 320 W/cm. The peak burnup reached by the carbide fuel is 127 GWd/t, without any fuel clad failure. The four sodium pumps have been operating satisfactorily for a cumulative time of more than 5,00,000 h. Creep, fatigue and fluence govern the life of the nuclear systems. Because of the reduced power and temperature at which the reactor has so far been operated, there is little ageing of the nuclear systems. The life of the nuclear components is being monitored by periodic surveillance. Periodic assessment of the fluence seen by reactor components is being made. The conventional systems have been in service for the past 19 years. Civil structures are 25 years old. These have been maintained by periodic preventive maintenance and replacement / repair wherever required. This paper details the various ageing management practices in FBTR. (author)

A capital cost reduction study on the fast breeder reactor plant

International Nuclear Information System (INIS)

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

1991-01-01

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

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)

Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

International Nuclear Information System (INIS)

Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

2013-01-01

The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

Basic research using the 250 KW research reactor triga in Ljubljana, Yugoslavia

International Nuclear Information System (INIS)

Dimic, V.

1983-01-01

The 25 KW Triga Mark II reactor of J. 'Stefan Institute' was commissioned on May 1966. During the last two years, it has been operated for about 4200 hr/year. According to experience gained with the reactor, most of the cost of reactor operation will be earned through isotope production for local hospitals and industries, performing low cost applied experiments and organizing training courses. The rest was provided through the Research Communities of the Republic of Slovenia. The reactor has been operated for 15 years without major problems and many basic research programmes have been performed. The research is being conducted in the following mainfields: solid state physics, neutron dosimetry, neutron radiography and autoradiography, reactor physics, examination of nuclear fuel using gamma scanning, irradiation of semiconducting materials and neutron activation analysis. (A.J)

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

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

Computational methodology for the Oak Ridge Research Reactor (ORR) and Bulk Shielding Reactor (BSR): cross-section and validation. Volume 1

International Nuclear Information System (INIS)

Miller, L.F.; Williams, M.L.

1986-03-01

A neutronics library suitable for low-enrichment uranium (LEU) and high-enrichment uranium (HEU) fueled cores for both the Oak Ridge Research Reactor (ORR) and the Bulk Shielding Reactor (BSR) is documented herein. The library is obtained from version V of the Evaluated Nuclear Data File (ENDF/B-V) and contains 223 nuclides weighted over a variety of region-dependent neutron spectra. Self-shielding and zone-weighting effects are incorporated with 227-group calculations for several reactor-core configurations. Libraries are archived for both transport and diffusion theory seven-group calculations. Complete listings of processing details are included so that libraries with different specifications can be easily obtained. Results from validation calculations indicate that the neutronics libraries obtained from this effort are suitable for neutronics computations for the ORR and BSR. 12 refs., 5 figs., 15 tabs

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

International Nuclear Information System (INIS)

Essadki, A.H.; Gourich, B.; Vial, Ch.; Delmas, H.; Bennajah, M.

2009-01-01

Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm 2 , but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H 2 microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

Research on reactor physics using the Very High Temperature Reactor Critical Assembly (VHTRC)

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1988-01-01

The High Temperature Engineering Test Reactor (HTTR), of which the research and development are advanced by Japan Atomic Energy Research Institute, is planned to apply for the permission of installation in fiscal year 1988, and to start the construction in the latter half of fisical year 1989. As the duty of reactor physics research, the accuracy of the nuclear data is to be confirmed, the validity of the nuclear design techniques is to be inspected, and the nuclear safety of the HTTR core design is to be verified. Therefore, by using the VHTRC, the experimental data of the reactor physics quantities are acquired, such as critical mass, the reactivity worth of simulated control rods and burnable poison rods, the temperature factor of reactivity, power distribution and so on, and the experiment and analysis are advanced. The cores built up in the VHTRC so far were three kinds having different lattice forms and degrees of uranium enrichment. The calculated critical mass was smaller by 1-5 % than the measured values. As to the power distribution and the reactivity worth of burnable poison rods, the prospect of satisfying the required accuracy for the design of the HTTR core was obtained. The experiment using a new core having axially different enrichment degree is planned. (K.I.)

Burnable absorber for the PIK reactor

International Nuclear Information System (INIS)

Gostev, V.V.; Smolskii, S.L.; Tchmshkyan, D.V.; Zakharov, A.S.; Zvezdkin, V.S.; Konoplev, K.A.

1998-01-01

In the reactor PIK design a burnable absorber is not used and the cycle duration is limited by the rods weight. Designed cycle time is two weeks and seams to be not enough for the 100 MW power research reactor equipped by many neutron beams and experimental facilities. Relatively frequent reloading reduces the reactor time on full power and in this way increases the maintenance expenses. In the reactor core fuel elements well mastered by practice are used and its modification was not approved. We try to find the possibilities of installation in the core separate burnable elements to avoid poison of the fuel. It is possible to replace a part of the fuel elements by absorbers, since the fuel elements are relatively small (diameter 5.15mm, uranium 235 content 7.14g) and there are more then 3800 elements in the core. Nevertheless, replacing decreases the fuel burnup and its consumption. In the PIK fuel assembles a little part of the volume is occupied by the dumb elements to create a complete package of the assembles shroud, that is necessary in the hydraulic reasons. In the presented report the assessment of such a replacement is done. As a burnable material Gadolinium was selected. The measurements or the beginning of cycle were performed on the critical facility PIK. The burning calculation was confirmed by measurements on the 18MW reactor WWR-M. The results give the opportunity to twice the cycle duration. The proposed modification of the fuel assembles does not lead to alteration in the other reactor systems, but it touch the burned fuel reprocessing technology. (author)

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

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

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Study of plutonium recycling physics in light water reactors

International Nuclear Information System (INIS)

Reuss, Paul

1979-10-01

A stock of plutonium from the reprocessing of thermal neutron reactor fuel is likely to appear in the next few years. The use of this plutonium as fuel replacing 235 U in thermal reactors is probably more interesting than simple stock-piling storage: immobilization of a capital which moreover would deteriorate by radioactive decay of isotope 241 also fissile and present to an appreciable extend in plutonium from reprocessing (half-life 15 years); recycling, on the other hand, will supply energy without complete degradation of the stock for fast neutron reactor loads, the burned matter having been partially renewed by conversion; furthermore the use of plutonium will meet the needs created by a temporary pressure on the naturel and/or enriched uranium market. For these two reasons the recycling of plutonium in thermal neutron reactors is being considered seriously today. The present work is confined to neutronic aspects and centres mainly on pressurized water-moderated reactors, the most highly developed at present in France. Four aspects of the problem are examined: 1. the physics of a plutonium-recycling reactor special features of neutronic phenomena with respect to the 'conventional' scheme of the 235 U burning reactor; 2. calculation of a plutonium-recycling reactor: adaptation of standard methods; 3. qualification of these calculations from the viewpoint of both data and inevitable approximations; 4. the fuel cycle and particularly the equivalence of fissile matters [fr

Neutronic evaluation of annular fuel rods to assemblies 13 x 13, 14 x 14 and 15 x 15

International Nuclear Information System (INIS)

Silva, Raphael H.M.; Ramos, Mario C.; Velasquez, Carlos E.; Silva, Clarysson A.M. da; Pereira, Cláubia; Costa, Antonella L.

2017-01-01

Research and development in nuclear reactor field has been proposed a new concept of fuel rod such as annular shape. The design of the annular fuel rods allows the coolant flow through the inner and outer side of it. Such project was proposed as an alternative to the traditional fuel rods used in LWR reactors. This new geometry allows an increase in power density in the reactor core with greater heat transfer from the fuel to the coolant which reduces the temperature in central region of the rod, in which a better configuration and dimension of fuel elements are aimed due to improvement of cooling in possible replacement of PWR traditional rods for annular rods. The aim of this work is to evaluate the neutronic parameters of fuel element with annular fuel rods where three configurations were studied: 13 x 13, 14 x 14 and 15 x 15. The goal is compare the neutronic between the advanced and the standard fuel assembly 16 x 16. In these studies, the external dimension and the moderator to fuel volume ratio (V M /V F ) of standard 16 x 16 is the same in all annular fuels assemblies. The MCNPX 2.6.0 (Monte Carlo N-Particle eXtended – version 2.6.0) code was used in all simulations. After all procedures, the annular fuel assemblies 13 have obtained greater neutronics parameters and were selected to more neutronics simulations. (author)

Neutronic evaluation of annular fuel rods to assemblies 13 x 13, 14 x 14 and 15 x 15

Energy Technology Data Exchange (ETDEWEB)

Silva, Raphael H.M.; Ramos, Mario C.; Velasquez, Carlos E.; Silva, Clarysson A.M. da; Pereira, Cláubia; Costa, Antonella L., E-mail: rapha.galo@hotmail.com, E-mail: marc5663@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Research and development in nuclear reactor field has been proposed a new concept of fuel rod such as annular shape. The design of the annular fuel rods allows the coolant flow through the inner and outer side of it. Such project was proposed as an alternative to the traditional fuel rods used in LWR reactors. This new geometry allows an increase in power density in the reactor core with greater heat transfer from the fuel to the coolant which reduces the temperature in central region of the rod, in which a better configuration and dimension of fuel elements are aimed due to improvement of cooling in possible replacement of PWR traditional rods for annular rods. The aim of this work is to evaluate the neutronic parameters of fuel element with annular fuel rods where three configurations were studied: 13 x 13, 14 x 14 and 15 x 15. The goal is compare the neutronic between the advanced and the standard fuel assembly 16 x 16. In these studies, the external dimension and the moderator to fuel volume ratio (V{sub M}/V{sub F}) of standard 16 x 16 is the same in all annular fuels assemblies. The MCNPX 2.6.0 (Monte Carlo N-Particle eXtended – version 2.6.0) code was used in all simulations. After all procedures, the annular fuel assemblies 13 have obtained greater neutronics parameters and were selected to more neutronics simulations. (author)

Verification test for radiation reduction effect and material integrity on PWR primary system by zinc injection

Energy Technology Data Exchange (ETDEWEB)

Kawakami, H.; Nagata, T.; Yamada, M. [Nuclear Power Engineering Corp. (Japan); Kasahara, K.; Tsuruta, T.; Nishimura, T. [Mitsubishi Heavy Industries, Ltd. (Japan); Ishigure, K. [Saitama Inst. of Tech. (Japan)

2002-07-01

Zinc injection is known to be an effective method for the reduction of radiation source in the primary water system of a PWR. There is a need to verify the effect of Zn injection operation on radiation source reduction and materials integrity of PWR primary circuit. In order to confirm the effectiveness of Zn injection, verification test as a national program sponsored by Ministry of Economy, Trade and Industry (METI) was started in 1995 for 7-year program, and will be finished by the end of March in 2002. This program consists of irradiation test and material integrity test. Irradiation test as an In-Pile-Test managed by AEAT Plc(UK) was performed using the LVR-15 reactor of NRI Rez in Check Republic. Furthermore, Out-of-Pile-Test using film adding unit was also performed to obtain supplemental data for In-Pile-Test at Takasago Engineering Laboratory of NUPEC. Material Integrity test was planned to perform constant load test, constant strain test and corrosion test at the same time using large scale Loop and slow strain extension rate testing (SSRT) at Takasago Engineering Laboratory of NUPEC. In this paper, the results of the verification test for Zinc program at present are discussed. (authors)

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

GoAmazon2014/15. Oxidation Flow Reactor Final Campaign Report

Energy Technology Data Exchange (ETDEWEB)

Jimenez, J. L. [Univ. of Colorado, Boulder, CO (United States); Day, D. A. [Univ. of Colorado, Boulder, CO (United States); Hu, W. [Univ. of Colorado, Boulder, CO (United States); Palm, B. B. [Univ. of Colorado, Boulder, CO (United States); Campuzano-Jost, P. [Univ. of Colorado, Boulder, CO (United States)

2016-03-01

The primary goal of the Green Ocean Amazon (GoAmazon2014/5) field campaign was to measure and mechanistically understand the formation of particle number and mass in a region affected by large tropical rainforest biogenic emissions and sometimes anthropogenic influence from a large urban center. As part of the two intensive operational periods (IOPs) and in collaboration with Pacific Northwest National Laboratory (PNNL) and Harvard, the Jimenez Group proposed to deploy a High-Resolution Time-of-Flight Aerosol Mass Spectrometer (HR-ToF-AMS), Thermal Denuder (TD), Scanning Mobility Particle Size (SMPS), two oxidation flow reactors (OFR; including supporting O₃, CO/CO₂/CH₄, RH analyzers), and a high volume filter sampler (MCV) for the measurement of gas and aerosol chemical, physicochemical, and volatility properties. The two IOPs were conducted during the wet season (February to March, 2014) and dry season (August to October, 2014). This proposal was part of a collaborative proposal involving other university and government laboratories.

Design study on small CANDLE reactor

Energy Technology Data Exchange (ETDEWEB)

Sekimoto, H; Yan, M [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

2007-07-01

A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burnup reactivity control mechanisms such as control rods are not required. This burnup strategy can derive many merits. The change of excess reactivity along burnup is theoretically zero, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Therefore, the operation of the reactor becomes much easier than the conventional reactors especially for high burnup reactors. The transportation and storage of replacing fuels become easy and safe, since they are free from criticality accidents. In our previous works it is appeared that application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. The average burnup of the spent fuel is about 40% that is equivalent to 40% utilization of the natural uranium without the reprocessing and enrichment. This reactor can be realized for large reactor, since the neutron leakage becomes small and its neutron economy becomes improved. In the present paper we try to design small CANDLE reactor whose performance is similar to the large reactor by increasing its fuel volume ration of the core, since its performance is strongly required for local area usage. Small long life reactor is required for some local areas. Such a characteristic that only natural uranium is required after second core is also strong merit for this case. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is

Design study on small CANDLE reactor

International Nuclear Information System (INIS)

Sekimoto, H.; Yan, M.

2007-01-01

A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burnup reactivity control mechanisms such as control rods are not required. This burnup strategy can derive many merits. The change of excess reactivity along burnup is theoretically zero, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Therefore, the operation of the reactor becomes much easier than the conventional reactors especially for high burnup reactors. The transportation and storage of replacing fuels become easy and safe, since they are free from criticality accidents. In our previous works it is appeared that application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. The average burnup of the spent fuel is about 40% that is equivalent to 40% utilization of the natural uranium without the reprocessing and enrichment. This reactor can be realized for large reactor, since the neutron leakage becomes small and its neutron economy becomes improved. In the present paper we try to design small CANDLE reactor whose performance is similar to the large reactor by increasing its fuel volume ration of the core, since its performance is strongly required for local area usage. Small long life reactor is required for some local areas. Such a characteristic that only natural uranium is required after second core is also strong merit for this case. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

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

Simulation evaluation of capacitor bank impact on increasing supply current for alumunium production

Science.gov (United States)

Hasan, S.; Badra, K.; Dinzi, R.; Suherman

2018-03-01

DC current supply to power the electrolysis process in producing aluminium at PT Indonesia Asahan Aluminium (Persero) is about 193 kA. At this condition, the load voltage regulator (LVR) transformer generates 0.89 lagging power factor. By adding the capacitor bank to reduce the harmonic distortion, it is expected that the supply current will increase. This paper evaluates capacitor bank installation impact on the system by using ETAP 12.0 simulation. It has been obtained that by installing 90 MVAR capacitor bank in the secondary part of LVR, the power factor is corrected about 8% and DC current increases about 13.5%.

A Study on the Kinetic Characteristics of Transmutation Process Reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae seon; Huh, Chang Wook; Kim, Doh Hyung [Seoul National University, Seoul (Korea, Republic of)

1997-07-01

The purpose of this study is to examine the transient heat transfer characteristics of liquid mental as the coolant used in accelerator-driven transmutation process reactor which is related the disposal of high-level radioactive nuclide. At current stage, the accelerator-driven transmutation process is investigated as the most appropriate method among many transmutation process methods. In this study, previous research works are investigated especially about the thermal hydraulics and kinetic behavior of coolant material including heat transfer of coolant in transmutation process reactor. A study on the heat transfer characteristics of liquid metal is performed based on the thermal hydraulic kinetic characteristics of liquid metal reactor which uses liquid metal coolant. Based on this study, the most appropriate material for the coolant of transmutation reactor will be recommended. 53 refs., 15 tabs., 33 figs. (author)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

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

Fuel assemblies for nuclear reactor

International Nuclear Information System (INIS)

Nishi, Akihito.

1987-01-01

Purpose: To control power-up rate at the initial burning stage of new fuel assemblies due to fuel exchange in a pressure tube type power reactor. Constitution: Burnable poisons are disposed to a most portion of fuel pellets in a fuel assembly to such a low concentration as the burn-up rate changes with time at the initial stage of the burning. The most portion means substantially more than one-half part of the pellets and gadolinia is used as burn-up poisons to be dispersed and the concentration is set to less than about 0.2 %. Upon elapse of about 15 days after the charging, the burnable poisons are eliminated and the infinite multiplication factors are about at 1.2 to attain a predetermined power state. Since the power-up rate of the nuclear reactor fuel assembly is about 0.1 % power/hour and the power-up rate of the fuel assembly around the exchanged channel is lower than that, it can be lowered sufficiently than the limit for the power-up rate practiced upon reactor start-up thereby enabling to replace fuels during power operation. (Horiuchi, T.)

Fast reactor knowledge preservation system: Taxonomy and basic requirements

International Nuclear Information System (INIS)

2008-01-01

The IAEA has taken the initiative to coordinate efforts of Member States in the preservation of knowledge in the area of fast reactors. In the framework of this initiative, the IAEA intends to create an international database compiling information from different Member States on fast reactors through a web portal. Other activities related to this initiative are being designed to accumulate and exchange information on the fast reactor area, to facilitate access to this information by users in different countries and to assist Member States in preserving the experience gained in their countries. The purpose of this publication is to develop a taxonomy of the Fast Reactor Knowledge Preservation System (FRKPS) that will facilitate the preservation of the world's fast reactor knowledge base, to identify basic requirements of this taxonomy on the basis of the experience gained in the fast reactor area, as well as results of previous IAEA activities on fast reactor knowledge preservation. The need for such taxonomy arises from the fact that during the past 15 years there has been stagnation in the development of fast reactors in the industrialized countries that were involved, earlier, in intensive development of this area. All studies on fast reactors have been stopped in countries such as Germany, Italy, the United Kingdom and the United States of America and the only work being carried out is related to the decommissioning of fast reactors. Many specialists who were involved in the studies and development work in this area in these countries have already retired or are close to retirement. In countries such as France, Japan and the Russian Federation that are still actively pursuing the evolution of fast reactor technology, the situation is aggravated by the lack of young scientists and engineers moving into this branch of nuclear power

Some Examples of Accident Analyses for RB Reactor

International Nuclear Information System (INIS)

Pesic, M.

2002-01-01

The RB reactor is heavy water critical assembly operated in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, since April 1959. The first Safety Analysis Report of the RB critical assembly was prepared in 1961/62. But, the first accidental analysis was done in late 1958 in aim the examine power transient and total equivalent doses received by the staff during the reactivity accident occurred on October 15, 1958. Since 1960, the RB reactor is modified few times. Beside initial natural uranium metal fuel rods, new fuel (TVR-S types) from 2% enriched metal uranium and 80% enriched UO 2 were available since 1962 and 1976, respectively. Also, modifications in control and safety systems of the reactor were done occasionally. Special reactor cores were created using all three types of fuel elements, among them, the coupled fast-thermal ones. Nuclear Safety Committee of the Vinca Institute, an independent regulatory body approved for usage all these modifications of the RB reactor. For those decisions of the Committee, the Preliminary Safety Analysis Reports were prepared that, beside proposed technical modifications and new regulation rules had included analyses of various possible accidents. Special attention is given and new methodology was proposed for thoroughly analyses of design based accidents related to coupled fast-thermal cores, that include reactor central zones filled by fuel elements without moderator. In these accidents, during assumed flooding of the fast zone by moderator, a very high reactivity could be inserted in the system with very high reactivity rate. It was necessary to provide that the safety system of the reactor had fast response to that accident and had enough high (negative) reactivity to shut down the reactor timely. In this paper, a brief overview of some accidents, methodology and computation tools used for the accident analyses at RB reactor are given. (author)

Reactor physics and reactor computations

International Nuclear Information System (INIS)

Ronen, Y.; Elias, E.

1994-01-01

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

Logistics of the research reactor fuel cycle: AREVA solutions

International Nuclear Information System (INIS)

Ohayon, David; Halle, Laurent; Naigeon, Philippe; Falgoux, Jean-Louis; Franck Obadia, Franck; Auziere, Philippe

2005-01-01

The AREVA Group Companies offer comprehensive solutions for the entire fuel cycle of Research Reactors comply with IAEA standards. CERCA and Cogema Logistics have developed a full partnership in the front end cycle. In the field of uranium CERCA and Cogema Logistics have the long term experience of the shipment from Russia, USA to the CERCA plant.. Since 1960, CERCA has manufactured over 300,000 fuel plates and 15,000 fuel elements of more than 70 designs. These fuel elements have been delivered to 40 research reactors in 20 countries. For the Back-End stage, Cogema and Cogema Logistics propose customised solutions and services for international shipments. Cogema Logistics has developed a new generation of packaging to meet the various needs and requirements of the Laboratories and Research Reactors all over the world, and complex regulatory framework. Comprehensive assistance dedicated, services, technical studies, packaging and transport systems are provided by AREVA for every step of research reactor fuel cycle. (author)

Heavy-Water Power Reactors. Proceedings Of A Symposium

International Nuclear Information System (INIS)

1968-01-01

Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

Heavy-Water Power Reactors. Proceedings Of A Symposium

Energy Technology Data Exchange (ETDEWEB)

NONE

1968-04-15

Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

Radiation effects on superconductivity in A15 materials

International Nuclear Information System (INIS)

Faehnle, M.

1981-01-01

At present the A15 superconductor Nb 3 Sn is one of the most attractive materials for the design of magnet systems for fusion reactors. There the materials are exposed to a high flux of neutrons up to 10 18 to 10 19 n/cm 2 during a continuous fusion reactor operation within ten years. As a result the critical parameters of the superconducting materials are changed which must be taken into account when designing reliable magnet systems. The neutron radiation damage in A15 materials thereby is characterized by small highly disordered regions within a less disordered matrix. The highly disordered regions are responsible for the increase of the critical current density after low-dose neutron irradiation of non-optimized materials and have an influence on the superconducting transition width. In contrast, the change of the superconducting parameters after high-dose irradiation may be understood essentially by considering the properties of the matrix alone. 23 refs

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Manipulator system for remote maintenance of fusion experimental reactor

International Nuclear Information System (INIS)

Shibanuma, Kiyoshi; Munakata, Tadashi; Murakami, Shin; Kondoh, Mitsunori.

1991-01-01

We have completed the conceptual design for a rail-mounted vehicle type remote maintenance system for the fusion experimental reactor (FER), which will be the first D-T burning reactor in Japan. We have fabricated a 1/5-scale model and confirmed the feasibility of the design. In this system, a rail is deployed into the vessel and supported at four horizontal ports. A vehicle then moves along the rail and handles in-vessel components with manipulators. The advantages of this concept are the high stiffness and high reliability of the rail, and the high mobility of the vehicle for efficient maintenance operations. In the FER, this concept is considered to be the first option for in-vessel maintenance. This paper describes the conceptual design of the system and the feasibility study using the 1/5-scale model. (author)

Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

Energy Technology Data Exchange (ETDEWEB)

Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

2015-06-15

Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

International Nuclear Information System (INIS)

Labib, Satira; King, Jeffrey

2015-01-01

Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort

Reactor

International Nuclear Information System (INIS)

Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

1979-01-01

Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

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

GE's advanced nuclear reactor designs

International Nuclear Information System (INIS)

Berglund, R.C.

1993-01-01

The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

1982-12-01

Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Plans for use of ENDF/B in reactor research in Indonesia

International Nuclear Information System (INIS)

Santoso, B.; Syaukat, A.; Subki, I.; Ganesan, S.

1989-07-01

Nuclear data are numerical constants of nature which quantify the nuclear behaviour of all elements and isotopes which make up the reactor medium and its environment, and which are needed as input for performing design calculations for safe and reliable operation of nuclear reactors. The nuclear data are available in the form of recommended values in specially formatted computerized files such as the Evaluated Nuclear Data File-B, known as ENDF/B. The development of base technology in the scheme of original reactor design calculations involves the mastering of the art of ENDF/B data processing. This paper briefly discusses the current status of this activity in Jakarta and gives an account of the future plans, with emphasis on the role of ENDF/B in reactor calculations. (author). 15 refs, 9 figs

Reactor container

International Nuclear Information System (INIS)

Kato, Masami; Nishio, Masahide.

1987-01-01

Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

Fast breeder reactor-block antiseismic design and verification

International Nuclear Information System (INIS)

Martelli, A.; Forni, M.

1988-02-01

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

Selection and challenges for LFR reactor materials

Energy Technology Data Exchange (ETDEWEB)

Weisenburger, A.; Jianu, A.; Del Giacco, M.; Fetzer, R.; Heinzel, A.; Mueller, G. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Pulsed Power and Microwave Technology

2013-07-01

Nuclear energy using Fast GenIV reactors can fulfil future demands concerning CO2 free, base load capability and sustainability. One of the most promising coolants especially due to its high thermal inertia is liquid lead (Pb). Since several years researches all over the world investigate this coolant and its impact on the reactor design and by that on the materials to be selected. The LEADER project, a follow up of ELSY, aims to design a prototypical demonstrator ALFRED and to continue with several design related aspects of the ELFR reactor. For a demonstrator the criteria of material selection are somewhat different to a commercial type like the ELFR. Material selection for ELFR of course considers all the aspects relevant for ALFRED plus the targeted burn up and the expected total dpa related damage especially of the fuel pins. In the past compatibility of structural material (steels like 316L, T91 and 15-15Ti (1.4970)) that can be employed for Pb cooled fast nuclear reactors were investigated in several EU projects like EUROTRANS and worldwide. Solubility of steel alloying elements like Ni, Fe, Cr is the driving force for the reduced corrosion resistance in contact with Pb. In-situ oxidation is the acknowledged measure to protect steels in Pb up to certain temperatures that are material dependent. Based on experiments and the derived temperature limits the average core outlet temperatures of ALFRED and the ELFR are set to 480 C. The most challenging conditions with respect to temperature are at the fuel assembly and the heat exchangers. For both, thin stable oxide scales with negligible reduction in heat transfer are the requested protection method. This presentation will give an overview on the selected materials for ALFRED and ELFR considering, beside pure compatibility, the influence of mechanical interaction like creep and fretting. (orig.)

BN-600 power unit 15-year operating experience

International Nuclear Information System (INIS)

Saraev, O.M.; Oshkanov, N.N.; Vylomov, V.V.

1996-01-01

Comprehensive experience has been gained with the operating fast reactor BN-600 with a power out of 600 MWe. This paper includes important performance results and gives also an overview of the experience gained from BN-600 NPP commercial operation during 15 years. (author). 2 figs, 1 tab

Special features of embrittlement of welded joints in shells of VVER-type reactors

International Nuclear Information System (INIS)

Kasatkin, O.G.

1999-01-01

At present, the atomic power engineering of Russia and Ukraine is based on water-water energy reactors of the VVER-440 and VVER-1000 type, with the electric power of 440 and 1000 MW, respectively. The majority of the VVER-440 reactors are installed in Russia, and VVER-1000 reactors operate in Ukraine. The reactors' shell (RS) is produced from cylindrical shells and a dished end welded together by circular joints under a flux. The RS of the VVER-440 reactor is produced from 15Kh3MFA steel, and the VVER-1000 reactors are produced from 15Kh2NMFA steel. The shell of the VVER-1000 reactor has an internal austenite coating. The condition of the RS metal is determined mainly by the critical brittleness temperature T b at which the impact toughness of specimens with a sharp notch reaches 60 J/cm 2 . The energy reactors, working in western countries, are characterised by a service life of 40 years and discussion is being carried out to extend this lifetime to 60 years. The design service life of the domestic reactors varies from 30 (RS VVER-440) to 40 (RS VVER-1000) years. According to investigations, the service life of the shells of these reactors is restricted by the properties of welded joints which are characterised by higher susceptibility to embrittlement than that of the parent metal, especially due to a higher content in the weld of phosphorus (RS VVER-440) or nickel (RS VVER-1000). Therefore, some experts believe that the actual service life of the RS is shorter than the design life. The accurate evaluation of the service life of welded joints in the RS is very important for the safety of service and also in the economic aspects, because the unjustified decrease of the permissible service life and premature shutdown of units of the nuclear power station result in huge losses

Effluents and releases of tritium from Novo-Voronezh-5 reactor

International Nuclear Information System (INIS)

Babenko, A.G.; Mekhedov, B.N.; Podporinova, L.E.; Popov, S.V.; Shalin, A.N.

1990-01-01

Results of systematic measurements of tritium concentration within technological systems of reactor of Novo-Voronezh NPP conducted to evaluate tritium effluents and releases and radiation doses to population from these effluents and releases are given. It is shown that 68% concerning tritium total amount were disposed into sewerage while 17% - through vent tube and 15% - with water and steam from secondary circuit systems. Standartized tritium effluents from WWER-1000 reactor for 5 year run constitute 15±1.9 GBq/MWxyear and it corresponds to mean value of effluents for foreign NPPs. Tritium concentration in the atmosphere constituted according to calculations (4.1-20)x10 -5 Bq/l. Conclusion is made about insignificant dose to population from tritium gaseous effluents. Detail study is necessary for dose connected with tritium contained in water effluents

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

Energy Technology Data Exchange (ETDEWEB)

Essadki, A.H., E-mail: essadki@hotmail.com [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Gourich, B. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Vial, Ch. [Laboratoire de Genie Chimique et Biochimique, LGCB-UBP/ENSCCF, 24 avenue des Landais, BP 206, 63174 Aubiere Cedex (France); Delmas, H. [Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France); Bennajah, M. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France)

2009-09-15

Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm{sup 2}, but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H{sub 2} microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Slurry reactors

Energy Technology Data Exchange (ETDEWEB)

Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

1979-08-01

Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-01-15

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

Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

International Nuclear Information System (INIS)

Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

2011-01-01

JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

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

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

A helium-cooled blanket design of the low aspect ratio reactor

International Nuclear Information System (INIS)

Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

1998-03-01

An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh

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)

Review of accident analyses of RB experimental reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-01-01

The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

Review of accident analyses of RB experimental reactor

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2003-01-01

Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

Space reactors. Progress report, October 1981-March 1982

International Nuclear Information System (INIS)

Ranken, W.A.

1983-01-01

Progress in design studies and technology for the SP-100 Project - successor to the Space Power Advanced Reactor (SPAR) Project - is reported for the period October 1, 1981 to March 31, 1982. The basis for selecting a high-temperature, UO 2 -fueled, heat-pipe-cooled reactor with a thermoelectric conversion system as the 100-kW/sub e/ reference design has been reviewed. Although no change has been made in the general concept, design studies have been done to investigate various reactor/conversion system coupling methods and core design modifications. Thermal and mechanical finite element modeling and three-dimensional Monte Carlo analysis of a core with individual finned fuel elements are reported. Studies of unrestrained fuel irradiation data are discussed that are relevant both to the core modeling work and to the design and fabrication of the first in-pile irradiation test, which is also reported. Work on lithium-filled core heat pipe development is described, including the attainment of 15.6 kW/sub t/ operation at 1525 K for a 2-m-long heat pipe with a 15.7-mm outside diameter. The successful operation of a 5.5-m-long, lightweight potassium/titanium heat pipe at 760 K is described, and test results of a thermoelectric module with GaP-modified SiGe thermoelectric elements are presented

Assessment value of 3-dimensional speckle tracking imaging for changes of early left ventricular longitudinal systolic function in patients with primary hypertension

Directory of Open Access Journals (Sweden)

Jing Yu

2016-08-01

Full Text Available Objective: To study the assessment value of 3-dimensional speckle tracking imaging for changes of early left ventricular longitudinal systolic function in patients with primary hypertension. Methods: Patients with primary hypertension who were treated in our hospital from May 2012 to October 2015 were selected, and 40 patients with left ventricular normal (LVN primary hypertension and 40 patients with left ventricular remodeling (LVR primary hypertension were screened according to Ganau typing and enrolled in the LVN group and LVR group of the study respectively; 40 cases of healthy volunteers who received physical examination in our hospital during the same period were selected as control group. Ultrasonic testing was conducted to determine conventional ultrasonic indicators and 3D-STI parameters, and serum was collected to determine AngII, ALD, TGF-β1 and Ang1-7 levels. Results: LVEDd, LVPWT and LVEF of LVN group were not significantly different from those of control group, LVEF of LVR group was not significantly different from those of LVN group and control group, and LVEDd and LVPWT of LVR group were significantly higher than those of LVN group and control group; absolute values of GLS, GCS, GRS and GAS as well as serum Ang1-7 level of LVN group was significantly lower than those of control group, serum AngII, ALD and TGF-β1 levels were higher than those of control group, absolute values of GLS, GCS, GRS and GAS as well as serum Ang1-7 level of LVR group was significantly lower than those of LVN group and control group, and serum AngII, ALD and TGF-β1 levels were higher than those of LVN group and control group; absolute values of GLS, GCS, GRS and GAS were negatively correlated with serum AngII, ALD and TGF-β1 levels, and positively correlated with serum Ang1-7 level. Conclusion: 3-dimensional speckle tracking imaging can be used for early evaluation of left ventricular longitudinal systolic function in patients with primary

Noble gas binary mixtures for gas-cooled reactor power plants

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

1993-05-13

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

Memory Deficit is Associated with Worse Functional Trajectories Among Older Adults in Low Vision Rehabilitation for Macular Disease

Science.gov (United States)

Whitson, Heather E.; Whitaker, Diane; Sanders, Linda L.; Potter, Guy G.; Cousins, Scott W.; Ansah, Deidra; McConnell, Eleanor; Pieper, Carl F.; Landerman, Lawrence; Steffens, David C.; Cohen, Harvey J.

2012-01-01

Objectives Older adults with macular disease are at increased risk of memory decline and incident dementia. Low vision rehabilitation (LVR) aims to preserve independence in people with irreversible vision loss, but comorbid memory problems could limit the success of rehabilitation. This study examined whether performance on a brief memory test is related to functional outcomes among older patients undergoing LVR for macular disease. Design Observational cohort study of patients receiving outpatient LVR Setting Academic center Participants 91 seniors (average age 80.1 years) with macular disease Measurements Memory was assessed at baseline with a 10-word list; memory deficit was defined as immediate recall of ≤ two words. Vision-related function was measured with the 25-item Visual Function Questionnaire (VFQ-25)administered at baseline and during subsequent interviews (mean length of follow up = 115 days). Linear mixed models (LMMs) were constructed to compare average trajectories of four VFQ-25 subscales: near activities, distance activities, dependency, and role difficulty. Results The 29.7% of patients with memory deficit tended to decline in ability to accomplish activities that involve near vision. Controlling for age, sex, and education, the functional trajectory of participants with memory deficit differed significantly from that of participants with better memory (p=0.002), who tended to report improvements in ability to accomplish near activities. Conclusion Among older adults receiving LVR for macular disease, those with memory deficit experienced worse functional trajectories in their ability to perform specific visually mediated tasks. A brief memory screen may help explain variability in rehabilitation outcomes and identify patients who might require special accommodations. PMID:23126548

The encapsulated nuclear heat source reactor for proliferation-resistant nuclear energy

International Nuclear Information System (INIS)

Brown, N.W.; Hossain, Q.; Carelli, M.D.; Conway, L.; Dzodzo, M.; Greenspan, E.; Saphier, D.

2001-01-01

The encapsulated nuclear heat source (ENHS) is a modular reactor that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor concept. It is a fast neutron spectrum reactor cooled by Pb-Bi using natural circulation. It is designed for passive load following, for high level of passive safety, and for 15 years without refueling. One of the unique features of the ENHS is that the fission-generated heat is transferred from the primary coolant to the secondary coolant across the reactor vessel wall by conduction-providing for an essentially sealed module that is easy to install and replace. Because the fuel is encapsulated within a heavy steel container throughout its life it provides a unique improvement to the proliferation resistance of the nuclear fuel cycle. This paper presents the innovative technology of the ENHS. (author)

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Space-time reactor kinetics for heterogeneous reactor structure

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1969-11-15

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

Update on Nonsurgical Lung Volume Reduction Procedures

Directory of Open Access Journals (Sweden)

J. Alberto Neder

2016-01-01

Full Text Available There has been a surge of interest in endoscopic lung volume reduction (ELVR strategies for advanced COPD. Valve implants, coil implants, biological LVR (BioLVR, bronchial thermal vapour ablation, and airway stents are used to induce lung deflation with the ultimate goal of improving respiratory mechanics and chronic dyspnea. Patients presenting with severe air trapping (e.g., inspiratory capacity/total lung capacity (TLC 225% predicted and thoracic hyperinflation (TLC > 150% predicted have the greatest potential to derive benefit from ELVR procedures. Pre-LVRS or ELVR assessment should ideally include cardiological evaluation, high resolution CT scan, ventilation and perfusion scintigraphy, full pulmonary function tests, and cardiopulmonary exercise testing. ELVR procedures are currently available in selected Canadian research centers as part of ethically approved clinical trials. If a decision is made to offer an ELVR procedure, one-way valves are the first option in the presence of complete lobar exclusion and no significant collateral ventilation. When the fissure is not complete, when collateral ventilation is evident in heterogeneous emphysema or when emphysema is homogeneous, coil implants or BioLVR (in that order are the next logical alternatives.

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Safety of light water reactors. Risks of nuclear technology

International Nuclear Information System (INIS)

Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

2012-01-01

The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

Conceptual design tool development for a Pb-Bi cooled reactor

International Nuclear Information System (INIS)

Lee, K. G.; Chang, S. H.; No, H. C.; Chunm, M. H.

2000-01-01

Conceptual design is generally ill-structured and mysterious problem solving. This leads the experienced experts to be still responsible for the most of synthesis and analysis task, which are not amenable to logical formulations in design problems. Especially because a novel reactor such as a Pb-Bi cooled reactor is going on a conceptual design stage, it will be very meaningful to develop the conceptual design tool. This tool consists of system design module with artificial intelligence, scaling module, and validation module. System design decides the optimal structure and the layout of a Pb-Bi cooled reactor, using design synthesis part and design analysis part. The designed system is scaled to be optimal with desired power level, and then the design basis accidents (Dbase) are analyzed in validation module. Design synthesis part contains the specific data for reactor components and the general data for a Pb-Bi cooled reactor. Design analysis part contains several design constraints for formulation and solution of a design problem. In addition, designer's intention may be externalized through emphasis on design requirements. For the purpose of demonstration, the conceptual design tool is applied to a Pb-Bi cooled reactor with 125 M Wth of power level. The Pb-Bi cooled reactor is a novel reactor concept in which the fission-generated heat is transferred from the primary coolant to the secondary coolant through a reactor vessel wall of a novel design. The Pb-Bi cooled reactor is to deliver 125 M Wth per module for 15 effective full power years without any on-site fuel handling. The conceptual design tool investigated the feasibility of a Pb-Bi cooled reactor. Application of the conceptual design tool will be, in detail, presented in the full paper. (author)

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

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

2013-01-01

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

Comparative performance of UASB and anaerobic hybrid reactors for the treatment of complex phenolic wastewater.

Science.gov (United States)

Ramakrishnan, Anushuya; Surampalli, Rao Y

2012-11-01

The performance of an upflow anaerobic sludge blanket (UASB) reactor and an anaerobic hybrid reactor (AHR) was investigated for the treatment of simulated coal wastewater containing toxic phenolics at different hydraulic retention times (0.75-0.33d). Fast start-up and granulation of biomass could be achieved in an AHR (45d) than UASB (58d) reactor. Reduction of HRT from 1.5 to 0.33d resulted in a decline in phenolics removal efficiency from 99% to 77% in AHR and 95% to 68% in UASB reactor respectively. AHR could withstand 2.5 times the selected phenolics loading compared to UASB reactor that could not withstand even 1.2 times the selected phenolics loading. Residence time distribution (RTD) study revealed a plug flow regime in the AHR and completely mixed regime in UASB reactor respectively. Energy economics of the reactors revealed that 12,159MJd(-1) more energy can be generated using AHR than UASB reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

RA Reactor

International Nuclear Information System (INIS)

1978-02-01

In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

Material and geometry options and performance characteristics for a test reactor

International Nuclear Information System (INIS)

Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

1993-01-01

For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

H Reactor

Data.gov (United States)

Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

Instrumentation for the advanced high-flux reactor workshop: proceedings

International Nuclear Information System (INIS)

Moon, R.M.

1984-01-01

The purpose of the Workshop on Instrumentation for the Advanced High-Flux Reactor, held on May 30, 1984, at the Oak Ridge National Laborattory, was two-fold: to announce to the scientific community that ORNL has begun a serious effort to design and construct the world's best research reactor, and to solicit help from the scientific community in planning the experimental facilities for this reactor. There were 93 participants at the workshop. We are grateful to the visiting scientists for their enthusiasm and interest in the reactor project. Our goal is to produce a reactor with a peak thermal flux in a large D 2 O reflector of 5 x 10 15 n/cm 2 s. This would allow the installation of unsurpassed facilities for neutron beam research. At the same time, the design will provide facilities for isotope production and materials irradiation which are significantly improved over those now available at ORNL. This workshop focussed on neutron beam facilities; the input from the isotope and materials irradiation communities will be solicited separately. The reactor project enjoys the full support of ORNL management; the present activities are financed by a grant of $663,000 from the Director's R and D Fund. However, we realize that the success of the project, both in realization and in use of the reactor, depends on the support and imagination of a broad segment of the scientific community. This is more a national project than an ORNL project. The reactor would be operated as a national user facility, open to any research proposal with high scientific merit. It is therefore important that we maintain a continuing dialogue with outside scientists who will be the eventual users of the reactor and the neutron beam facilities. The workshop was the first step in establishing this dialogue; we anticipate further workshops as the project continues

Innovative Energy Planning and Nuclear Option Using CANDLE Reactors

Energy Technology Data Exchange (ETDEWEB)

Sekimoto, H; Nagata, A; Mingyu, Y [Tokyo Institute of Technology, Tokyo (Japan)

2008-07-01

A new reactor burn-up strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burn-up strategy can derive many merits. The change of excess reactivity along burn-up is theoretically zero for ideal equilibrium condition, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed during life of operation. Therefore, the operation of the reactor becomes much easier than the conventional reactors. The infinite-medium neutron multiplication factor of replacing fuel is less than unity. Therefore, the transportation and storage of replacing fuels becomes easy and safe, since they are free from criticality accidents. Small long life fast reactor with CANDLE burn-up concept has investigated with depleted uranium as a replacing fuel. Both core diameter and height are chosen to be 2.0 m, and the thermal power is 200 MW. Lead-bismuth is used as a coolant, and nitride (enriched N-15) fuel are employed. The velocity of burning region along burn-up is less than 1.0 cm/year that enables a long life design easily. The core averaged discharged fuel burn-up is about 40 percent. It is about ten times of light water reactor burn-up. The spent fuel volume becomes one-tenth of light water reactor spent fuel. If a light water reactor with a certain power output has been operated for 40 years, the CANDLE reactor can be operated for 2000 years with the same power output and with only depleted uranium left after fuel production for the light water reactor. The system does not need any reprocessing or enrichment. Therefore, the reactor operation becomes very safe, the waste

Innovative Energy Planning and Nuclear Option Using CANDLE Reactors

International Nuclear Information System (INIS)

Sekimoto, H.; Nagata, A.; Mingyu, Y.

2008-01-01

A new reactor burn-up strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burn-up strategy can derive many merits. The change of excess reactivity along burn-up is theoretically zero for ideal equilibrium condition, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed during life of operation. Therefore, the operation of the reactor becomes much easier than the conventional reactors. The infinite-medium neutron multiplication factor of replacing fuel is less than unity. Therefore, the transportation and storage of replacing fuels becomes easy and safe, since they are free from criticality accidents. Small long life fast reactor with CANDLE burn-up concept has investigated with depleted uranium as a replacing fuel. Both core diameter and height are chosen to be 2.0 m, and the thermal power is 200 MW. Lead-bismuth is used as a coolant, and nitride (enriched N-15) fuel are employed. The velocity of burning region along burn-up is less than 1.0 cm/year that enables a long life design easily. The core averaged discharged fuel burn-up is about 40 percent. It is about ten times of light water reactor burn-up. The spent fuel volume becomes one-tenth of light water reactor spent fuel. If a light water reactor with a certain power output has been operated for 40 years, the CANDLE reactor can be operated for 2000 years with the same power output and with only depleted uranium left after fuel production for the light water reactor. The system does not need any reprocessing or enrichment. Therefore, the reactor operation becomes very safe, the waste

Reactor container

International Nuclear Information System (INIS)

Naruse, Yoshihiro.

1990-01-01

The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

Hybrid reactors

International Nuclear Information System (INIS)

Moir, R.W.

1980-01-01

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

2003-10-01

Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

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

Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

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

Multiregion reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

Measurements and preliminary interpretation of K-Reactor foundation response to man-made seismic excitation

International Nuclear Information System (INIS)

Lee, R.C.; Stephenson, D.E.

1992-01-01

In support of the Savannah River Technology Center (SRTC) effort to develop K-Reactor seismic design basis ground motions, SRTC monitored local high-explosive tests at a ''free-field'' site adjacent to K-Reactor and on the -40 level on the foundation of K-Reactor. The high-explosive tests were part of the SRTC/United States Geological Survey (USGS) regional refraction and attenuation experiment that used deeply buried high explosive charges near New Ellenton, Snelling, and at more distant South Carolina sites. The primary purpose of the Reactor measurements are to compare the relative amplitude and frequency content of ambient noise and shot generated ground motions measured at the K-Reactor foundation level and in the ''free-field'' so that foundation effects to ground motions can be documented and possibly incorporated in the facility design basis. Data analysis indicates that one of the five high explosive tests provided sufficient excitations at K-Reactor to produce satisfactory signal-to-noise between about 1 Hz and 15 Hz. Within this frequency band, Fourier spectral amplitude ratios of motions recorded within the first 10 seconds of first motion show substantial reductions (30 endash 50%) on shot radial and transverse components for frequencies greater than about 3 to 5 Hz. Approximately 50% reductions between 10 to 15 Hz were seen on vertical component ratios, and amplifications of 100% at 4 Hz and 5 Hz endash 6 Hz

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

Science.gov (United States)

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Reactor building

International Nuclear Information System (INIS)

Maruyama, Toru; Murata, Ritsuko.

1996-01-01

In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Global results concerning the operation of the reactors at the Grenoble nuclear research centre

International Nuclear Information System (INIS)

Jacquemain, M.; Marouby, R.

1964-01-01

The Grenoble Nuclear Research Centre has 3 reactors of the open-core swimming-pool type: Melusine (2 MW) operating since 1959, Siloe (15 MW) operating since 1963, Siloette (100 kW) operating since 1964 The report describes the operating conditions of these reactors and the improvements which have been made to increase the flux in the irradiation rigs and to increase the safety and the regularity of operation. The advantages are also explained of having on the same site, close to one another, several reactors with wide ranges of flux. (authors) [fr

Reactor physics challenges in GEN-IV reactor design

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

2005-02-15

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

Reactor physics challenges in GEN-IV reactor design

International Nuclear Information System (INIS)

Driscoll, Michael K.; Hejzlar, Pavel

2005-01-01

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

Reactor Vessel Surveillance Program for Advanced Reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

2008-10-15

This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

Design and selection of materials for sodium cooled fast reactors

International Nuclear Information System (INIS)

Chetal, S.C.

2011-01-01

Sodium cooled fast reactors are currently in operation, under construction or under design by a number of countries. The design of sodium cooled fast reactor is covered by French RCC - MR code and ASME code NH. The codes cover rules as regards to materials, design and construction. These codes do not cover the effect of irradiation and environment. Elevated temperature design criteria in nuclear codes are much stringent in comparison to non nuclear codes. Sodium corrosion is not an issue in selection of materials provided oxygen impurity in sodium is controlled for which excellent reactor operating experience is available. Austenitic stainless steels have remained the choice for the permanent structures of primary sodium system. Stabilized austenitic stainless steel are rejected because of poor operating experience and non inclusion in the design codes. Route for improved creep behaviour lies in compositional modifications in 316 class steel. However, the weldability needs to be ensured. For cold leg component is non creep regime, SS 304 class steel is favoured from overall economics. Enhanced fuel burn up can be realized by the use of 9-12%Cr 1%Mo class steel for the wrapper of MOX fuel design, and cladding and wrapper for metal fuel reactors. Minor compositional modifications of 20% cold worked 15Cr-15Ni class austenitic stainless steel will be a strong candidate for the cladding of MOX fuel design in the short term. Long term objective for the cladding will be to develop oxide dispersion strengthened steel. 9%Cr 1%Mo class steel (Gr 91) is an ideal choice for integrated once through sodium heated steam generators. One needs to incorporate operating experience from reactors and thermal power stations, industrial capability and R and D feedback in preparing the technical specifications for procurement of wrought products and welding consumables to ensure reliable operation of the components and systems over the design life. The paper highlights the design approach

A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2015-01-01

In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe

A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [retired, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Royal Canadian Navy, Ottawa, Ontario (Canada)

2015-03-15

In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe

Sixth meeting of the International Working Group on Water Reactor Fuel Performance and Technology, Vienna, 14-15 May 1984

International Nuclear Information System (INIS)

1984-10-01

The summary report of the 6th meeting of the IWGFPT had been approved. The review of IWGFPT activities since the last meeting are given in Appendix 3, which is a list of meetings and proceedings publications for that period. Co-ordinated research programmes were discussed in the fields of fuel element cladding interaction with water coolant in power reactors and examination and documentation methodology for water reactor fuel. The main activities for the next three years were settled. The next Working Group meeting will be held at the VIC, November 1985

Physics design of advanced steady-state tokamak reactor A-SSTR2

International Nuclear Information System (INIS)

Nishio, Satoshi; Ushigusa, Kenkichi

2000-10-01

Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

Low temperature oxidation of hydrocarbons using an electrochemical reactor

DEFF Research Database (Denmark)

Ippolito, Davide

conversion was a complex function of multiple variables: the microstructure of the backbone, the polarization resistance of the electrodes, both at OCV and under polarization, the electrical and morphological properties of the infiltrated material and the specific reaction conditions like the propene......This study investigated the use of a ceramic porous electrochemical reactor for the deep oxidation of propene. Two electrode composites, La0.85Sr0.15MnO3±d/Ce0.9Gd0.1O1.95 (LSM/CGO) and La0.85Sr0.15FeMnO3/Ce0.9Gd0.1O1.95 (LSF/CGO), were produced in a 5 single cells stacked configuration and used...... prolonged polarization was able to partially counteract the instability of the infiltrated Ce0.9Gd0.1O1.95. This project demonstrated the possibility to enhance the oxidation of propene by polarization in a porous ceramic reactor. The infiltration of different active materials helped to increase...

Application of vanadium alloys to a fusion reactor blanket

Energy Technology Data Exchange (ETDEWEB)

Bethin, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center)

1984-05-01

Vanadium and vanadium alloys are of interest in fusion reactor blanket applications due to their low induced radioactivity and outstanding elevated temperature mechanical properties during neutron irradiation. The major limitation to the use of vanadium is its sensitivity to oxygen impurities in the blanket environment, leading to oxygen embrittlement. A quantitative analysis was performed of the interaction of gaseous impurities in a helium coolant with vanadium and the V-15Cr-5Ti alloy under conditions expected in a fusion reactor blanket. It was shown that the use of unalloyed V would impose severe restrictions on the helium gas cleanup system due to excessive oxygen buildup and embrittlement of the metal. However, internal oxidation effects and the possibly lower terminal oxygen solubility in the alloy would impose much less severe cleanup constraints. It is suggested that V-15Cr-5Ti is a promising candidate for certain blanket applications and deserves further consideration.

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Enrichment of 15N and 10B isotopes by chemical exchange process

International Nuclear Information System (INIS)

D'Souza, A.B.; Sonwalkar, A.S.; Subrahmanyam, B.V.; Valladares, B.A.

1994-01-01

Many processes are available for separation of stable isotopes like distillation, chemical exchange, thermal diffusion, gaseous diffusion, centrifuge etc. Chemical exchange process is eminently suitable for separation of isotopes of light elements. Work done on separation and enrichment of two of the stable isotopes viz. 15 N and 10 B in Chemical Engineering Division is presented. 15 N is widely used as a tracer in agricultural research and 10 B is used in nuclear industry as control rod material, soluble reactor poison, neutron detector etc. The work on 15 N isotope resulted in a pilot plant, which was the only source of this material in the country for many years and later it was translated into a production plant as M/s. RCF Ltd. The work done on the ion-exchange process for enrichment of 10 B isotope which is basically a chemical exchange process, is now being updated into a pilot plant to produce enriched 10 B to be used as soluble reactor poison. (author)

Startup of Torrey Pines Mark III and Puerto Rico Nuclear Center reactors with TRIGA-FLIP fuel

Energy Technology Data Exchange (ETDEWEB)

Chesworth, R. H. [Gulf E and ES, San Diego, CA (United States)

1972-07-01

This paper discusses the characteristics of TRIGA FLIP cores in two different geometries: the normal TRIGA single-rod geometry as typified by the installation in the Torrey Pines Mark III reactor; and the four-rod cluster geometry as typified by the conversion core installed in the Puerto Rico Nuclear Center reactor at Mayaguez. In both reactors the fuel is 8-1/2 wt % uranium, 70% enriched in U-235. The hydrogen to zirconium atom ratio is 1.5 to 1.65 and the cladding material is stainless steel. The basic neutronic characteristics of the fuel in both reactor installations are briefly discussed.

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Nuclear Reactor Physics

Science.gov (United States)

Stacey, Weston M.

2001-02-01

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

Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

International Nuclear Information System (INIS)

Eustache, S.

1999-01-01

This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

Variations in the chemical speciation behaviour of radioiodines in the Tarapur Boiling Water Reactor

International Nuclear Information System (INIS)

Venkateswaran, G.; Gokhale, A.S.; Moorthy, P.N.

1998-01-01

The chemical behaviour of radioiodines in the primary coolant of the Tarapur Boiling Water Reactor has been studied under different operating conditions. During normal operation, radioiodines speciated mainly as I - (≅60%) and IO 3 - (≅35%) with 2 . At 1-5 h into reactor shutdown conditions, radioiodines existed predominantly as IO 3 - species (>80%). Beyond 5 h after shutdown, quantitative conversion of IO 3 - to I - was observed to occur in about 20 h duration. Long time after reactor shutdown, radioiodines were present in the coolant as I - species only. A quantitative conversion of near carrier-free IO 3 - to I - was observed in laboratory low dose rate (0.95 kGy/h), low and high dose gamma irradiation experiments in near neutral solutions both in absence and presence of externally added H 2 O 2 . However, near carrier-free I - solutions irradiated under the same conditions yielded ≅15% IO 3 - species only which is in agreement with the literature data. The radioiodine speciation behaviour in reactor water has been explained by a qualitative model coupling iodine release from defective fuel elements and the associated gamma irradiation effects. (author)

Importance of helical pitch parameter in LHD-type heliotron reactor designs

International Nuclear Information System (INIS)

Goto, T.; Suzuki, Y.; Yanagi, N.; Watanabe, K.Y.; Imagawa, S.; Sagara, A.

2010-11-01

In the design studies of the LHD-type heliotron reactors, one of the key issues is to secure sufficient blanket spaces. In this respect, helical pitch parameter γ is quite important because it significantly affects both the coil and plasma shapes. In order to understand the effect of helical pitch parameter on the design window quantitatively, a system design code for the LHD-type heliotron reactors has been developed and parametric scans were carried out with 3 cases of γ=1.15, 1.20 and 1.25. It becomes clear that the possible design window of heliotron reactors strongly depends on the engineering constraints: stored magnetic energy of coil system, blanket space, and neutron wall load. γ=1.20 is optimum from the viewpoint of moderating the physics requirements, but γ=1.15 has a robustness to the change in the physics and engineering conditions. Since the design windows are quite sensitive to the engineering constraints and physics conditions, the further detailed study on design feasibility of advanced engineering components and the effect of γ on the physics conditions is expected to optimize the value of γ. (author)