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Sample records for start-up fission reactor

  1. Research nuclear reactor start-up simulator

    International Nuclear Information System (INIS)

    Sofo Haro, M.; Cantero, P.

    2009-01-01

    This work presents the design and FPGA implementation of a research nuclear reactor start-up simulator. Its aim is to generate a set of signals that allow replacing the neutron detector for stimulated signals, to feed the measurement electronic of the start-up channels, to check its operation, together with the start-up security logic. The simulator presented can be configured on three independent channels and adjust the shape of the output pulses. Furthermore, each channel can be configured in 'rate' mode, where you can specify the growth rate of the pulse frequency in %/s. Result and details of the implementation on FPGA of the different functional blocks are given. (author)

  2. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  3. Start-up simulations of the PULSAR pulsed tokamak reactor

    International Nuclear Information System (INIS)

    Werley, K.A.; Bathke, C.G.

    1993-01-01

    Start-up conditions are examined for a pulsed tokamak reactor that uses only inductively driven plasma current (and bootstrap current). A zero-dimensional (profile-averaged) model containing plasma power and particle balance equations is used to study several aspects of plasma start-up, including: (1) optimization of the start-up pathway; (2) tradeoffs of auxiliary start-up heating power versus start-up time; (3) volt-second consumption; (4) thermal stability of the operating point; (5) estimates of the diverter heat flux and temperature during the start-up transient; (6) the sensitivity of the available operating space to allowed values of the H confinement factor; and (7) partial-power operations

  4. Reactor water quality degradation suppressing method upon reactor start up

    International Nuclear Information System (INIS)

    Maeda, Katsuharu.

    1993-01-01

    Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

  5. Program of RA reactor start-up to nominal power

    International Nuclear Information System (INIS)

    1959-01-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is described in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members

  6. Power start up of the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    After accomplishing the physical start-up of the reactor, the power start-up was carried out in February 1984. The power of the reactor has reached: 10 KW on 6/2/1984, 100 KW on 7/2/1984, 200 KW and 300 KW on 8/2/1984; 400 KW and nominal power 500 KW on 9/2/1984. The reactivity temperature coefficient and the xenon poisoning were determined. 3 figs., 12 tabs

  7. Control device for start-up of reactor depressurization system

    International Nuclear Information System (INIS)

    Suzuki, Hiroshi; Saito, Minoru; Oda, Shingo; Miura, Satoshi; Hashimoto, Koji; Tate, Hitoshi; Fujii, Kazunobu

    1998-01-01

    The present invention concerns are emergency reactor core cooling system (ECCS) of a BWR type reactor and provides a control device for start-up of an automatic depressurization system. Namely, the device has an object of preventing erroneous opening of a main steam escape safety value when testing a start-up signal circuit of an automatic depressurization system for testing the automatic depressurization system. A start-up signal circuit receives both signals of a reactor container pressure high signal and a reactor pressure vessel water level low signal and outputs an automatic start-up signal for compulsorily opening a main steam escape safety valve automatically. A test switch having a self-holding circuit is disposed to a central control chamber. A test signal circuit is disposed for preventing transfer of an erroneous start-up signal to the main steam escape safety valve due to a simulation signal during output test signals by the test switch. (I.S.)

  8. Fusion reactor start-up without an external tritium source

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, S., E-mail: Shanliang.Zheng@ccfe.ac.uk; King, D.B.; Garzotti, L.; Surrey, E.; Todd, T.N.

    2016-02-15

    Highlights: • Investigated the feasibility (including plasma physics, neutronics and economics) of starting a fusion reactor from running pure D–D fusion reactor to gradually move towards the D–T operation. • Proposed building up tritium from making use of neutrons generated by D–D fusion reactions. • Studied plasma physics feasibility for pure D–D operation and provided consistent fusion power and neutron yield in the plasma with different mixture of deuterium and tritium. • Discussed the economics aspect for operating a pure D–D fusion reactor towards a full-power D–T fusion reactor. - Abstract: It has long been recognised that the shortage of external tritium sources for fusion reactors using D–T, the most promising fusion fuel, requires all such fusion power plants (FPP) to breed their own tritium. It is also recognised that the initial start-up of a fusion reactor will require several kilograms of tritium within a scenario in which radioactive decay, ITER and subsequent demonstrator reactors are expected to have consumed most of the known tritium stockpile. To circumvent this tritium fuel shortage and ultimately achieve steady-state operation for a FPP, it is essential to first accumulate sufficient tritium to compensate for loss due to decay and significant retention in the materials in order to start a new FPP. In this work, we propose to accumulate tritium starting from D–D fusion reactions, since D exists naturally in water, and to gradually build up the D–T plasma targeted in fusion reactor designs. There are two likely D–D fusion reaction channels, (1) D + D → T + p, and (2) D + D → He3 + n. The tritium can be generated via the reaction channel ‘(1)’ and the 2.45 MeV neutrons from ‘(2)’ react with lithium-6 in the breeding blanket to produce more tritium to be fed back into plasma fuel. Quantitative evaluations are conducted for two blanket concepts to assess the feasibility and suitability of this approach to FPP

  9. UASB reactor start up in real scale for malting effluent

    International Nuclear Information System (INIS)

    Lopez, I.; Passeggi, M.; Boix, C.; Barcia, R.; Borzacconi, L.; Liebermann, L.

    2005-01-01

    An Imhoff tank was reconstructed into a 250 m3 UASB reactor in order to treat a malting plant wastewater.The UASB was inoculated with sludge from an anaerobic lagoon used for slaughterhouse wastewater treatment.The fat present in the inoculated sludge did not affect the start up performance. After two month of operation the reactor achieved full load with COD removal higher than 80% and a biogas production of 300 m3/day. The mean diameter of granules was 0a.milimeters and the Specific Methanogenic Activity was 0.25 g COD/gVSS.d The banquet development throughout time (solids concentrations at different heights, granule size, methanogenic activity) was monitored. A yield coefficient of 0.09 gVSS/g COD rem was found

  10. Automatic start-up system of nuclear reactor based on sequence control technology

    International Nuclear Information System (INIS)

    Zhang Yao; Zhang Dafa; Peng Huaqing

    2009-01-01

    A conceptive design of an automatic start-up system based on the sequence control for the nuclear reactors is given in this paper, so as to solve the problems during the start-up process, such as the long operation time, low automatic control level and high accident rate. The start-up process and its requirements are analyzed in detail at first. Then,the principle, the architecture, the key technologies of the automatic start-up system of nuclear reactors are designed and discussed. With the designed system, the automatic start-up of the nuclear reactor can be realized,the work load of the operator can be reduced,and the safety and efficiency of the nuclear power plant during its start-up can be improved. (authors)

  11. Temelin-1 reactor unit commissioning and start-up

    International Nuclear Information System (INIS)

    Palecek, K.

    2002-01-01

    The Temelin-1 commissioning process was affected substantially by the change in the Czech political situation in 1989. The effects thereof were both favourable and unfavourable. Among favourable effects are the replacement of the original Instrumentation and Control System by a more advanced system and design changes which have brought about additional improvement of the Temelin NPP design safety, although on the other hand, this had an adverse impact on the time span and price of the power plant construction. Additional adverse effects included an unstable political and economic situation, associated with frequent changes in the management of the utility CEZ, a.s. (owner of the plant) as well as frequent replacement of persons in the position of the managing director of the Temelin plant itself. Despite all the difficulties encountered, Temelin-1 reactor unit could be ultimately put into trial operation in June 2002. (author)

  12. First physical start-up for the first pulsed reactor in China

    International Nuclear Information System (INIS)

    Huang Wenlou; Tan Rilin; Xie Yuqi; Chai Songshan; Li Yingfa; He Qianming; Zhou Bin

    1993-01-01

    The characteristics and the test results of initial loading fuel and first physical start-up for the first pulsed reactor in China (PRC-1) are described. Safe measure to ensure safety of first physical start-up are also described. The experiments show that performances of PRC-1 are in accord with design requirements

  13. Use of coaxial plasma guns to start up field-reversed-mirror reactors

    International Nuclear Information System (INIS)

    Smith, A.C. Jr.; Carlson, G.A.; Eddleman, J.L.; Hartman, C.W.; Neef, W.S. Jr.

    1980-01-01

    Application of a magnetized coaxial plasma gun for start-up of a field-reversed-mirror reactor is considered. The design is based on preliminary scaling laws and is compared to the design of the start-up gun used in the Beta II experiment

  14. New start-up channels and multichannel analyzer at the RB reactor; Novi start-up kanali i videkanalni analizator na reaktoru Rb

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Markovic, H; Vranic, S; Dimitrijevic, Z; Pesic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1978-01-15

    New start-up channels and a multichannel analyzer were purchased in 1977 for the RB reactor. Both start-up channels contain BF{sub 3} neutron detectors, preamplifier, amplifier, single-channel analyzer, scaler, ratemeter, control unit, recording instrument. This document contains detailed technical description of these devices as well as characteristics of the multichannel analyzer which is being tested and will be used for measuring irradiation in the vicinity of the reactor.

  15. Design of data sampler in intelligent physical start-up system for nuclear reactor

    International Nuclear Information System (INIS)

    Wang Yinli; Ling Qiu

    2007-01-01

    It introduces the design of data sampler in intelligent physical start-up system for nuclear reactor. The hardware frame taking STμPSD3234A as the core and the firmware design based on USB interface are discussed. (authors)

  16. Study and application of digital physical start-up system for nuclear reactor

    International Nuclear Information System (INIS)

    Qu Ronghong; Li Baoxiang; Xu Xiaolin

    2004-01-01

    The digital physical start-up system for nuclear reactor is introduced. The system was used successfully in physical start-up experiment of 10 MW high-temperature gas-cooled reactor. It is proved practically that the system not only runs reliably and calculates both rapidly and correctly and relieves the loads of operators, but also has the better characters of monitoring and showing the real-time results of experiments than the analog systems. (author)

  17. Start-up analysis of INET-5 MW district heating prototype reactor

    International Nuclear Information System (INIS)

    Li Tianshu

    1991-09-01

    The main features and thermohydraulic design parameters of the INET-5 MW reactor (INET: Institute of Nuclear Technology of Tsinghua University, Beijing) are presented. The start-up process and the effect of thermohydraulic instability on start-up process have been analyzed. The main obstacle of start-up process of INET-5 MW reactor is to pass the instability region from 1 atm to normal operation condition. For avoiding instability, the start-up process should be divided into two steps. The results of three different start-up proposals calculated by DACOL code are given and compared. The possibility of instabilities for each proposal has been checked. The checked results show that there is no instability during start-up of the three proposals. So, it is supposed that the INET-5 MW reactor can safely and stably reach the operation conditions. Finally, some conclusions about the effect of instability on start-up in boiling mode of INET-5MW reactor are given

  18. Start-up test of the prototype heavy water reactor 'FUGEN', (1)

    International Nuclear Information System (INIS)

    Ando, Hideki; Kawahara, Toshio

    1982-01-01

    The advanced thermal prototype reactor ''Fugen'' is a heavy water-moderated, boiling light water-cooled power reactor with electric output of 165 MW, which has been developed since 1966 as a national project. The start-up test was begun in March, 1978, being scheduled for about one year, and in March, 1979, it passed the final pre-use inspection and began the full scale operation. In this paper, the result of the start-up test of Fugen is reported. From the experience of the start-up test of Fugen, the following matters are important for the execution of start-up test. 1) Exact testing plan and work schedule, 2) the organization to perform the test, 3) the rapid evaluation of test results and the reflection to next testing plan, and 4) the reflection of test results to rated operation, regular inspection and so on. In the testing plan, the core characteristics peculiar to Fugen, and the features of heavy water-helium system, control system and other equipment were added to the contents of the start-up test of BWRs. The items of the start-up test were reactor physics test, plant equipment performance test, plant dynamic characteristic test, chemical and radiation measurement, and combined test. The organization to perform the start-up test, and the progress and the results of the test are reported. (Kako, I.)

  19. Device for the nuclear reactor automatic start-up and power control

    International Nuclear Information System (INIS)

    Nikiforov, B.N.; Volkov, A.V.; Ogon'kov, A.I.

    1978-01-01

    A description and flowsheet of a reactor start-up and power-control automatic device containing no nonlinear elements with a relay characteristic are presented. The device consists of two independent channels for measuring the physical power and time (period) constant of the reactor. Requirements for the device are considered, based on the condition of a minimum permissible number of a servomechanism operations due to fluctuations of an input signal which appear because of the statistical nature of processes taking place in the reactor. It is noted that the threshold amplifier used in the device allows a considerable decrease of the reactor start-up time

  20. An analysis of reactivity prediction during the reactor start-up process

    International Nuclear Information System (INIS)

    Bajgl, Josef; Krysl, Vaclav; Svarny, Jiri

    2015-01-01

    The different VVER-440 core fuel loadings subcriticality evaluations are performed during the start-up process by boron dilution or control assembly withdrawn by macrocode MOBY-DICK calculations. The dynamic reactivity and quasicritical reactivity are compared and sensitivity of reactivity prediction at the low boundary of start-up interval (ρ = -0,01) has been provided on the basis of different modelling of ionization chamber (IC) response calculation. Special attention is paid to the impact of power distribution and spontaneous fission distribution form factor on IC response correction during control assembly movement. Precision and robustness of different corrections of IC signal processing in real core start-up processed IC signals was evaluated.

  1. Optimal relations of the parameters ensuring safety during reactor start-up

    International Nuclear Information System (INIS)

    Yurkevich, G.P.

    2004-01-01

    Procedure and equations for the determination of optimal ratio between parameters allowing safe removal of reactor in critical state are suggested. Initial pulse frequency of pulsed start-up channel and power of neutron source are decreased by reduced rate of changing reactivity during automatic start-up, disposition of pulsed neutron detector in the range with neutron flux density to 5·10 12 s -1 cm -2 at standard power, separate signal of period for the use in chains of automatic start-up and emergency protection, reduction of pulses frequency of the start-up channel (the frequency is equal to 4000 c -1 ). Procedure and equations for the determination of optimal parameters are effected with the account of statistic character of pulsed detector frequency and false outlet signal [ru

  2. Development of intelligent physical start-up system for nuclear reactor

    International Nuclear Information System (INIS)

    Wang Canhui; Li Xiang; Huang Liyuan; Fu Guoen; Hu Hai

    2008-01-01

    In this paper, the Intelligent physical start-up system for nuclear reactor introduced the system composing, hardware design and software design. The system has some merits such as handy operation, fast and accurate mathematic and nicer human-machine interface. (authors)

  3. Analysis of the start-up and control of a particle bed reactor

    International Nuclear Information System (INIS)

    Lazareth, O.W.; Araj, K.J.; Horn, F.L.; Ludewig, H.; Powell, J.R.

    1987-01-01

    This study describes the modeling of start-up transients in Particle Bed Reactors (PBR) for burst electric power. Two computer programs have been developed to analyze the start-up process. The first program (named KINETIC) analyzes the entire fuel element, calculating time dependent solutions for power and the temperature distribution in the packed bed. The second program (named SPHEAT, for Spherical Heating) calculates time-dependent temperatures inside individual, cladded fuel particles. The two programs provide powerful analytical tools for evaluation of material and geometrical options, power and time constraints, and conditions that could lead to element failures

  4. On Stability of Natural-circulation-cooled Boiling Water Reactors during Start-up (Experimental Results)

    International Nuclear Information System (INIS)

    Manera, A.; Van der Hagen, T.H.J.J.

    2002-01-01

    The characteristics of flashing-induced instabilities, which are of importance during the start-up phase of natural-circulation Boiling Water Reactors (BWRs), are studied. Experiments at typical start-up conditions (low power and low pressure) are carried out on a steam/water natural circulation loop. The mechanism of flashing-induced instability is analyzed in detail and it is found that non-equilibrium between phases and enthalpy transport plays an important role in the instability process. Pressure and steam volume in the steam dome are found to have a stabilizing effect. The main characteristics of the instabilities have been analyzed. (authors)

  5. Microbiology and performance of a methanogenic biofilm reactor during the start-up period.

    Science.gov (United States)

    Cresson, R; Dabert, P; Bernet, N

    2009-03-01

    To understand the interactions between anaerobic biofilm development and process performances during the start-up period of methanogenic biofilm reactor. Two methanogenic inverse turbulent bed reactors have been started and monitored for 81 days. Biofilm development (adhesion, growth, population dynamic) and characteristics (biodiversity, structure) were investigated using molecular tools (PCR-SSCP, FISH-CSLM). Identification of the dominant populations, in relation to process performances and to the present knowledge of their metabolic activities, was used to propose a global scheme of the degradation routes involved. The inoculum, which determines the microbial species present in the biofilm influences bioreactor performances during the start-up period. FISH observations revealed a homogeneous distribution of the Archaea and bacterial populations inside the biofilm. This study points out the link between biodiversity, functional stability and methanogenic process performances during start-up of anaerobic biofilm reactor. It shows that inoculum and substrate composition greatly influence biodiversity, physiology and structure of the biofilm. The combination of molecular techniques associated to a biochemical engineering approach is useful to get relevant information on the microbiology of a methanogenic growing biofilm, in relation with the start-up of the process.

  6. Impact of reactor configuration on anammox process start-up: MBR versus SBR.

    Science.gov (United States)

    Tao, Yu; Gao, Da-Wen; Fu, Yuan; Wu, Wei-Min; Ren, Nan-Qi

    2012-01-01

    Anaerobic ammonium oxidation (anammox) is an energy saving biological nitrogen removal process which was limited to slow growth rate of anammox bacteria during start-up period. This study investigated the start-up of anammox process by a laboratory sequential batch reactor (SBR) for 218 days and subsequently modified the reactor as a membrane bioreactor (MBR) for 178 days. Modification of a SBR as MBR with installation of an external membrane module resulted in acceleration of specific anammox activity by 19 times. The acceleration of specific anammox activity with MBR was further confirmed by starting-up another MBR for a 242 day period. Molecular microbial analyses showed that Candidatus "Brocadia anammoxidans" and Candidatus "Kuenenia stuttgartiensis" were the dominant species in the inocula and biomass developed in the reactor. The start-up with MBR appeared to be more effective than SBR for the enrichment of anammox bacteria due to high sludge retention property of MBR configuration. Copyright © 2011 Elsevier Ltd. All rights reserved.

  7. Start-up and bacterial community compositions of partial nitrification in moving bed biofilm reactor.

    Science.gov (United States)

    Liu, Tao; Mao, Yan-Jun; Shi, Yan-Ping; Quan, Xie

    2017-03-01

    Partial nitrification (PN) has been considered as one of the promising processes for pretreatment of ammonium-rich wastewater. In this study, a kind of novel carriers with enhanced hydrophilicity and electrophilicity was implemented in a moving bed biofilm reactor (MBBR) to start up PN process. Results indicated that biofilm formation rate was higher on modified carriers. In comparison with the reactor filled with traditional carriers (start-up period of 21 days), it took only 14 days to start up PN successfully with ammonia removal efficiency and nitrite accumulation rate of 90 and 91%, respectively, in the reactor filled with modified carriers. Evident changes of spatial distributions and community structures had been detected during the start-up. Free-floating cells existed in planktonic sludge, while these microorganisms trended to form flocs in the biofilm. High-throughput pyrosequencing results indicated that Nitrosomonas was the predominant ammonia-oxidizing bacterium (AOB) in the PN system, while Comamonas might also play a vital role for nitrogen oxidation. Additionally, some other bacteria such as Ferruginibacter, Ottowia, Saprospiraceae, and Rhizobacter were selected to establish stable footholds. This study would be potentially significant for better understanding the microbial features and developing efficient strategies accordingly for MBBR-based PN operation.

  8. Simulation on the start-up of MED seawater desalination system coupled with nuclear heating reactor

    International Nuclear Information System (INIS)

    Ge Zhihua; Du Xiaoze; Yang Lijun; Yang Yongping; Wu Shaorong

    2008-01-01

    The mathematical control model for dynamic start-up process of the VTE-MED seawater desalination system was established employing the previous developed non-linear differential equations for system design and performance analysis. The influences on the start-up process of the operating parameters, such as the initial feed brine flow rate and the top brine temperature were analyzed. The relationships among the feed brine flow rate, the gained output ratio (GOR) and the start-up time were also investigated, which can be evidence to determine the optimal initial feed brine flow rate. The results also indicate that the system can consume the total heat rating generated by the low temperature nuclear heating reactor (LT-NHR) even at the most initial start-up stage, implying the present desalination system has excellent coupling characteristics with the LT-NHR. With necessary experiments verifications, the start-up control model developed in this paper can be the theoretical base for the analysis of dynamic performances of the seawater desalination system

  9. Successful hydraulic strategies to start up OLAND sequencing batch reactors at lab scale.

    Science.gov (United States)

    Schaubroeck, Thomas; Bagchi, Samik; De Clippeleir, Haydée; Carballa, Marta; Verstraete, Willy; Vlaeminck, Siegfried E

    2012-05-01

    Oxygen-limited autotrophic nitrification/denitrification (OLAND) is a one-stage combination of partial nitritation and anammox, which can have a challenging process start-up. In this study, start-up strategies were tested for sequencing batch reactors (SBR), varying hydraulic parameters, i.e. volumetric exchange ratio (VER) and feeding regime, and salinity. Two sequential tests with two parallel SBR were performed, and stable removal rates > 0.4 g N l(-1) day(-1) with minimal nitrite and nitrate accumulation were considered a successful start-up. SBR A and B were operated at 50% VER with 3 g NaCl l(-1) in the influent, and the influent was fed over 8% and 82% of the cycle time respectively. SBR B started up in 24 days, but SBR A achieved no start-up in 39 days. SBR C and D were fed over 65% of the cycle time at 25% VER, and salt was added only to the influent of SBR D (5 g NaCl l(-1)). Start-up of both SBR C and D was successful in 9 and 32 days respectively. Reactor D developed a higher proportion of small aggregates (0.10-0.25 mm), with a high nitritation to anammox rate ratio, likely the cause of the observed nitrite accumulation. The latter was overcome by temporarily including an anoxic period at the end of the reaction phase. All systems achieved granulation and similar biomass-specific nitrogen removal rates (141-220 mg N g(-1) VSS day(-1)). FISH revealed a close juxtapositioning of aerobic and anoxic ammonium-oxidizing bacteria (AerAOB and AnAOB), also in small aggregates. DGGE showed that AerAOB communities had a lower evenness than Planctomycetes communities. A higher richness of the latter seemed to be correlated with better reactor performance. Overall, the fast start-up of SBR B, C and D suggests that stable hydraulic conditions are beneficial for OLAND while increased salinity at the tested levels is not needed for good reactor performance. © 2012 The Authors. Microbial Biotechnology © 2012 Society for Applied Microbiology and Blackwell Publishing

  10. Prediction of the stability of BWR reactors during the start-up process

    International Nuclear Information System (INIS)

    Ruiz E, J.A.; Castillo D, R.; Blazquez M, J.B.

    2004-01-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  11. Procedure of determination of the nuclear reactor start-up power

    International Nuclear Information System (INIS)

    Brandt, K.D.; Griese, K.; Guehne, F.

    1985-01-01

    The invention has been aimed at a determination of the thermal reactor power during the start-up period and the commissioning resp. preferably under an extremely low thermal power in a range from 0.001 to 1%. In addition to it also the power range up to 100% shall be covered. The external gamma-ray flux density as a function of the thermal reactor power is measured in several overlapping partial measuring ranges. A suitable measuring device transforms the input signals into an electrically measured quantity proportional to the reactor power

  12. Application of the fractional neutron point kinetic equation: Start-up of a nuclear reactor

    International Nuclear Information System (INIS)

    Polo-Labarrios, M.-A.; Espinosa-Paredes, G.

    2012-01-01

    Highlights: ► Neutron density behavior at reactor start up with fractional neutron point kinetics. ► There is a relaxation time associated with a rapid variation in the neutron flux. ► Physical interpretation of the fractional order is related with non-Fickian effects. ► Effect of the anomalous diffusion coefficient and the relaxation time is analyzed. ► Neutron density is related with speed and duration of the control rods lifting. - Abstract: In this paper we present the behavior of the variation of neutron density when the nuclear reactor power is increased using the fractional neutron point kinetic (FNPK) equation with a single-group of delayed neutron precursor. It is considered that there is a relaxation time associated with a rapid variation in the neutron flux and its physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. We analyzed the case of increase the nuclear reactor power when reactor is cold start-up which is a process of inserting reactivity by lifting control rods discontinuously. The results show that for short time scales of the start-up the neutronic density behavior with FNPK shows sub-diffusive effects whose absorption are government by control rods velocity. For large times scale, the results shows that the classical equation of the neutron point kinetics over predicted the neutron density regarding to FNPK.

  13. A novel start-up procedure for natural-circulation boiling water reactors

    International Nuclear Information System (INIS)

    Annalisa Manera; Frank Schaefer

    2005-01-01

    Full text of publication follows: The elimination of recirculation pumps and associated systems, as proposed for natural-circulation Boiling Water Reactors (BWRs), allow a great simplification in the design of BWRs. On the other hand, it has been shown both experimentally and analytically that such a new reactor configuration makes the system susceptible to thermal-hydraulic instabilities during the start-up phase (so-called flashing-induced instabilities). Therefore, appropriate start-up procedures have to be planned to avoid instabilities in natural-circulation BWRs. Not many proposals of start-up procedures for natural-circulation BWRs are reported in literature, but all authors agree on the fact that the system should be pressurized before the transition to two-phase circulation is allowed. Nayak [1] and Jiang and coauthors [2] proposed to externally pressurize the system by injecting in the pressure vessel respectively steam produced in a separate boiler or nitrogen. Once the pressure in the reactor vessel is high enough, the reactor power can be increased to achieve two-phase natural circulation. Unfortunately, the procedure suggested by Nayak requires an external boiler of adequate volume and power and the related connecting piping to the reactor vessel, while the procedure suggested by Jiang and coauthors requires an additional system for the nitrogen storage and the related connecting piping to the reactor vessel. The external pressurization does not accomplish to the requirements of simplicity that are at the very base of natural circulation BWRs design and it is thus not recommendable. Cheung and Rao [3] suggested a start-up procedure in which the reactor is first filled with water at 80 deg. C at a pressure of 0.55 bar. The reactor is made critical and is pressurized in conditions of single-phase circulation up to a pressure of 63 bar. At this pressure a sudden transition to two-phase operation is achieved by opening the MSIVs (Main Steam Isolation

  14. Effect of calcium on microbial aggregation during UASB reactor start-up

    Energy Technology Data Exchange (ETDEWEB)

    Mahoney, E.M.; Varangu, L.K.; Cairns, W.L.; Kosaric, N.; Murray, R.G.E.

    1987-01-01

    The dynamics of granule formation were studied using cells from two bench-scale UASB reactors. The objective was to elucidate factors which influence formation and maintenance of highly active self-agglomerated microbial biomass. Simultaneous examination of biological and physical parameters was performed during the start-up of a calcium-positive (100 mg/l) reactor and a reactor without added calcium. The influence of carbon nutrients and Ca++ on the cell surface and microbial aggregation was studied. The granules formed in both reactors but were larger in the calcium-positive reactor in which they settled 3-4 times faster. A higher rate of biomass accumulation also was evident in the calcium-positive reactor and this allowed a more frequent increase in the substrate loading rate and earlier development of the granular sludge. (Refs. 17).

  15. Reactor core flow measurements during plant start-up using non-intrusive flow meter CROSSFLOW

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, V.; Sharp, B.; Gurevich, A., E-mail: vkanda@amag-inc.com, E-mail: bsharp@amag-inc.com, E-mail: agurevich@amag-inc.com [Advanced Measurement & Analysis Group Inc., Ontario (Canada); Gurevich, Y., E-mail: yuri.gurevich@daystartech.ca [Daystar Technologies Inc., Ontario (Canada); Selvaratnarajah, S.; Lopez, A., E-mail: sselvaratnarajah@amag-inc.com, E-mail: alopez@amag-inc.com [Advanced Measurement & Analysis Group Inc., Ontario (Canada)

    2013-07-01

    For the first time, direct measurements of the total reactor coolant flow and the flow distribution between the inner reactor zone and the outer zone were conducted using the non-intrusive clamp on ultrasonic cross-correlation flow meter, CROSSFLOW, developed and manufactured by Advanced Measurement & Analysis Group Inc. (AMAG). The measurements were performed at Bruce Power A Unit 1 on the Pump Discharge piping of the Primary Heat Transport (PHT) system during start-up. This paper describes installation processes, hydraulic testing, uncertainty analysis and traceability of the measurements to certified standards. (author)

  16. Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

    International Nuclear Information System (INIS)

    Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

    1980-01-01

    Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

  17. Determination of reactor parameters during start up test at the Taiwan NPP, Unit 1

    International Nuclear Information System (INIS)

    Astakhov, S.; Kravchenko, A.; Kraynov, Ju.; Nasedkin, A.; Tsyganov, S.

    2006-01-01

    Unit 1 of Taiwan NPP with WWER-1000 reactor reached the first criticality at December 20 of 2005. Series of start up experiments were carried out under scientific advisory of RRC 'Kurchatov Institute' specialists. At the Hot Zero Power state the reactivity coefficients, control rod group and scram worth were measured and symmetry of the core loading and power reactivity effect due to reaching 1% of nominal were assessed. Paper describes in brief special features of experiments and presented some results obtained at a measurements (Authors)

  18. Fusion core start-up, ignition, and burn simulations of reversed-field pinch (RFP) reactors

    International Nuclear Information System (INIS)

    Chu, Y.Y.

    1988-01-01

    A transient reactor simulation model is developed to investigate and simulate the start-up, ignition, and burn of a reversed-field pinch reactor. The simulation is based upon a spatially averaged plasma balance model with field profiles obtained from MHD quasi-equilibrium analysis. Alpha particle heating is estimated from Fokker-Planck calculations. The instantaneous plasma current is derived from a self-consistent circuit analysis for plasma/coil/eddy current interactions. The simulation code is applied to the TITAN RFP reactor design which features a compact, high-power-density reversed-field pinch fusion system. A contour analysis is performed using the steady-state global plasma balance. The results are presented with contours of constant plasma current. A saddle point is identified in the contour plot which determined the minimum value of plasma current required to achieve ignition. In the simulations of the TITAN RFP reactor, the OH-driven super-conducting EF coils are found to deviate from the required equilibrium values as the induced plasma current increases. A set of basic results from the simulation of TITAN RFP reactor yield a picture of RFP plasma operation in a reactor. Investigations of eddy currents are also presented and have very important in reactor design

  19. Analytic method study of point-reactor kinetic equation when cold start-up

    International Nuclear Information System (INIS)

    Zhang Fan; Chen Wenzhen; Gui Xuewen

    2008-01-01

    The reactor cold start-up is a process of inserting reactivity by lifting control rod discontinuously. Inserting too much reactivity will cause short-period and may cause an overpressure accident in the primary loop. It is therefore very important to understand the rule of neutron density variation and to find out the relationships among the speed of lifting control rod, and the duration and speed of neutron density response. It is also helpful for the operators to grasp the rule in order to avoid a start-up accident. This paper starts with one-group delayed neutron point-reactor kinetics equations and provides their analytic solution when reactivity is introduced by lifting control rods discontinuously. The analytic expression is validated by comparison with practical data. It is shown that the analytic solution agrees well with numerical solution. Using this analytical solution, the relationships among neutron density response with the speed of lifting control rod and its duration are also studied. By comparing the results with those under the condition of step inserted reactivity, useful conclusions are drawn

  20. Reactor Start-up and Control Methodologies: Consideration of the Space Radiation Environment

    International Nuclear Information System (INIS)

    Bragg-Sitton, Shannon M.; Holloway, James Paul

    2004-01-01

    The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable the accomplishment of ambitious space exploration missions. The natural radiation environment in space provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Initial investigation using MCNPX 2.5.b for proton transport through the SAFE-400 reactor indicates a secondary neutron net current of 1.4x107 n/s at the core-reflector interface, with an incoming current of 3.4x106 n/s due to neutrons produced in the Be reflector alone. This neutron population could provide a reliable startup source for a space reactor. Additionally, this source must be considered in developing a reliable control strategy during reactor startup, steady-state operation, and power transients. An autonomous control system is developed and analyzed for application during reactor startup, accounting for fluctuations in the radiation environment that result from changes in vehicle location (altitude, latitude, position in solar system) or due to temporal variations in the radiation field, as may occur in the case of solar flares. One proposed application of a nuclear electric propulsion vehicle is in a tour of the Jovian system, where the time required for communication to Earth is significant. Hence, it is important that a reactor control system be designed with feedback mechanisms to automatically adjust to changes in reactor temperatures, power levels, etc., maintaining nominal operation without user intervention. This paper will evaluate the potential use of secondary neutrons produced by proton interactions in the reactor vessel as a startup source for a space reactor and will present a

  1. Experimental investigation on flow behavior during start-up of a heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin; Zhang Youjie

    1997-09-01

    An experimental simulation study on the transition from pressurized to boiling operation of a low-temperature, natural circulation nuclear heating reactor (5 MW) developed by INET of Tsinghua University is presented. The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The manifestation of different kinds of two-phase flow instabilities, namely geyser instability, flashing instability and low-steam quality density wave instability on the transition from pressurized to boiling operation is described. The mechanism of flashing instability, which has never been studied well on this field, is especially interpreted. It is suggested that the start-up process, from initial condition to boiling operation condition, should consist of three steps: (1) increasing of initial pressure by means of a noncondensable gas (N 2 ), which is a very effective method to eliminate geyser instability and flashing instability at lower pressure. (2)start-up of the reactor at this pressurized condition with a constant heat flux under the limited value of q = 0.15 MW·m -2 , which controls the exit temperature of the heated section below the one of net vapor generation, the low steam quality density wave oscillation can be avoided. (3) transition to a lower pressure, boiling operation. The method of transition with low-heat flux and low-inlet subcooling is proposed: at pressurized operation condition, by reducing the heat flux to its lowest level, releasing the noncondensable gas and increasing the heat flux gradually (dq/dt -2 ·min -1 ), during which the low-steam quality density wave oscillation can be prevented from occurring, then the boiling operation condition can be achieved through adjusting the heat flux and inlet subcooling to their designed value. A stable transition from pressurized to boiling operation of the 5 MW reactor is achieved by careful selection of the thermohydraulic parameters. (7 refs., 7 figs., 1

  2. Physical start up of the Dalat nuclear research reactor with the core configuration exempt from neutron trap

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    The nominal power of the reconstructed Dalat reactor is of 500 KW. After a meticulous preparation the Russian and Vietnamese teams have proceeded to the physical reactor start-up in November 1983 with the core configuration exempt from the neutron trap. The reactor has reached the physical criticality at 19h50 on 1 November 1983. The report delineates different steps of the start-up procedure. 2 refs., 3 figs., 7 tabs

  3. Method of start-up of rotary plug sealing devices in FBR type reactors

    International Nuclear Information System (INIS)

    Sakuragi, Masanori; Akita, Haruo

    1980-01-01

    Purpose: To rapidly and safely start-up the rotary plug sealing device by controlling to eliminate the pressure difference in the pressures of gases exerting on the liquid surfaces in the inner and the outer cylinders of a sealing alloy vessel in the rotary plug of a FBR type reactor. Method: In a case where an abnormal state results in the pressure difference of gases exerted on the liquid surfaces in the inner and the outer cylinders of a vessel charged with sealing alloy in a rotary plug and the sealing valve for the back-up gas supply tube is rapidly closed to seal the sealing portion, the pressure in the gas supply tube is controlled so that the pressure difference in the gases exerted on the liquid surfaces in the inner and outer cylinders while closing the sealing valve. Then, after conforming that the pressure is controlled to a predetermined level at which the pressure difference can be regarded to be zero, the sealing valve is gradually opened while regulating the pressure in the gas supply tube so as to maintain the pressure difference to a predetermined level. This prevents the occurrence of external disturbances upon opening of the sealing valve and enables rapid and safety start-up for the rotary plug sealing device. (Moriyama, K.)

  4. Simulations and field tests of a reactor coolant pump emergency start-up by means of remote gas units

    International Nuclear Information System (INIS)

    Omahen, P.; Gubina, F.

    1992-01-01

    The problem of the reactor coolant pump start-up in case of emergency by means of remote gas power plant units was analyzed. In this paper a simulation model is developed which enabled a detailed simulation of the transient process occurring at the start-up. The start-up of the RCP motor set was simulated in case of available one and two gas units. The field tests were performed and the measured variable values complied well with the simulation results. Two gas units have been determined as a safe start-up scheme of the RCP motor set considering for safety reasons accepted busbars and motor protection settings. A derived model for deep rotor bars was experimentally confirmed as effective means for the RCP motor set start-up transient simulation. Start-up procedures have been designed and adopted to the safety procedures of the Nuclear Power Plant Krsko

  5. Implementation of nuclear power plant simulation in start-up commissioning of reactor control system

    International Nuclear Information System (INIS)

    Yang Zongwei; Huang Tieming; Feng Guangyu; Luan Zhenhua; Lin Meng; Zhu Lizhi

    2009-01-01

    Based on the nuclear power thermal-hydraulic model, Labview graphical programming language and virtual instrument data acquisition technology, this paper describes a dedicate test platform to solve the problem that the reactor control system (RRC) can not be evaluated and analyzed far before the actual startup of the unit. By connecting the test platform to the nuclear Digital Control System (DCS), the step-by-step closed-looped test and global function test of RRC system were performed, the dynamic validation and logical function demonstration for RRC were realized, and a lot of configuration mistakes of RRC and nonconformity were solved. The test for unit 3 of Ling'ao phase II has proved that the implementation of nuclear power plant simulation in the start-up commissioning of RRC can greatly reduce the risk of normal power operation and great transient tests, with which the term of startup for overall unit test can be greatly shortened. (authors)

  6. Program of RA reactor start-up to nominal power; Program dizanja reaktora 'RA' na nominalnu snagu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is desed in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members.

  7. Power start up of the Dalat nuclear research reactor; Khoi dong nang luong lo phan ung hat nhan Da Lat

    Energy Technology Data Exchange (ETDEWEB)

    Hien, Pham Duy; Huy, Ngo Quang; Long, Vu Hai; Mai, Tran Khanh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    After accomplishing the physical start-up of the reactor, the power start-up was carried out in February 1984. The power of the reactor has reached: 10 KW on 6/2/1984, 100 KW on 7/2/1984, 200 KW and 300 KW on 8/2/1984; 400 KW and nominal power 500 KW on 9/2/1984. The reactivity temperature coefficient and the xenon poisoning were determined. 3 figs., 12 tabs.

  8. Baikal-1 stand complex. Preparation and carrying out of the first energy start-up of the IVG-1 reactor

    International Nuclear Information System (INIS)

    Tikhomirov, L.N.

    1995-01-01

    The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs

  9. Correction method for critical extrapolation of control-rods-rising during physical start-up of reactor

    International Nuclear Information System (INIS)

    Zhang Fan; Chen Wenzhen; Yu Lei

    2008-01-01

    During physical start-up of nuclear reactor, the curve got by lifting the con- trol rods to extrapolate to the critical state is often in protruding shape, by which the supercritical phenomena is led. In the paper, the reason why the curve was in protruding was analyzed. A correction method was introduced, and the calculations were carried out by the practical data used in a nuclear power plant. The results show that the correction method reverses the protruding shape of the extrapolating curve, and the risk of reactor supercritical phenomena can be reduced using the extrapolated curve got by the correction method during physical start-up of the reactor. (authors)

  10. HAC and fission reactors

    International Nuclear Information System (INIS)

    Fujiwara, I.; Moriyama, H.; Tachikawa, E.

    1984-01-01

    In the fission process, newly formed fission products undergo hot atom reactions due to their energetic recoil and abnormal positive charge. The hot atom reactions of the fission products are usually accompanied by secondary effects such as radiation damage, especially in condensed phase. For reactor safety it is valuable to know the chemical behaviour and the release behaviour of these radioactive fission products. Here, the authors study the chemical behaviour and the release behaviour of the fission products from the viewpoint of hot atom chemistry (HAC). They analyze the experimental results concerning fission product behaviour with the help of the theories in HAC and other neighboring fields such as radiation chemistry. (Auth.)

  11. Report of blind start-up experiments carried out on the reactor Cabri between 4. and 8. July 1966

    International Nuclear Information System (INIS)

    Filipczak, N.; Filipczak, W.; Furet, J.; Kaiser, J.

    1967-01-01

    The blind start-up of a reactor without any neutronic data concerning a relatively wide range of power dynamics can be necessary when difficulties arise in the positioning of the detector or in neutron-gamma discrimination near the multiplying medium. The object of the experiments carried out on the reactor Cabri was to check the very complete analysis of the start-up accident which was studied on an analogue computer. The number of experiments carried out (12) is not sufficient to allow a definite conclusion. Nevertheless the blind start-up method advocated by N. FILIPCZAK and W. FILIPCZAK does not appear to be incompatible with the security during the operational phase (on condition that its dynamic characteristics are close to that of the reactor Cabri). (authors) [fr

  12. An operation protocol for facilitating start-up of single-stage autotrophic nitrogen removing reactors based on process stoichiometry

    DEFF Research Database (Denmark)

    Mutlu, A. Gizem; Vangsgaard, Anna Katrine; Sin, Gürkan

    2012-01-01

    Start-up and operation of single-stage nitritation/anammox reactor employing complete autotrophic nitrogen can be difficult. Keeping the performance criteria and monitoring the microbial community composition may not be easy or fast enough to take action on time. In this study, a control strategy...

  13. Short-sludge age EBPR process – Microbial and biochemical process characterisation during reactor start-up and operation

    DEFF Research Database (Denmark)

    Valverde Pérez, Borja; Wágner, Dorottya Sarolta; Lóránt, Bálint

    2016-01-01

    . In this paper, we report the start-up and operation of a short-SRT enhanced biological phosphorus removal (EBPR) system operated as a sequencing batch reactor (SBR) fed with preclarified municipal wastewater, which is supplemented with propionate. The microbial community was analysed via 16S rRNA amplicon...

  14. Start-Up Characteristics of a Granule-Based Anammox UASB Reactor Seeded with Anaerobic Granular Sludge

    Directory of Open Access Journals (Sweden)

    Lei Xiong

    2013-01-01

    Full Text Available The granulation of anammox sludge plays an important role in the high nitrogen removal performance of the anammox reactor. In this study, anaerobic granular sludge was selected as the seeding sludge to start up anammox reactor in order to directly obtain anammox granules. Results showed that the anammox UASB reactor was successfully started up by inoculating anaerobic granular sludge, with substrate capacity of 4435.2 mg/(L·d and average ammonium and nitrite removal efficiency of 90.36% and 93.29%, respectively. During the start-up course, the granular sludge initially disintegrated and then reaggregated and turned red, suggesting the high anammox performance. Zn-Fe precipitation was observed on the surface of granules during the operation by SEM-EDS, which would impose inhibition to the anammox activity of the granules. Accordingly, it is suggested to relatively reduce the trace metals concentrations, of Fe and Zn in the conventional medium. The findings of this study are expected to be used for a shorter start-up and more stable operation of anammox system.

  15. Start-Up Characteristics of a Granule-Based Anammox UASB Reactor Seeded with Anaerobic Granular Sludge

    Science.gov (United States)

    Wang, Yun-Yan; Tang, Chong-Jian; Chai, Li-Yuan; Xu, Kang-Que; Song, Yu-Xia

    2013-01-01

    The granulation of anammox sludge plays an important role in the high nitrogen removal performance of the anammox reactor. In this study, anaerobic granular sludge was selected as the seeding sludge to start up anammox reactor in order to directly obtain anammox granules. Results showed that the anammox UASB reactor was successfully started up by inoculating anaerobic granular sludge, with substrate capacity of 4435.2 mg/(L·d) and average ammonium and nitrite removal efficiency of 90.36% and 93.29%, respectively. During the start-up course, the granular sludge initially disintegrated and then reaggregated and turned red, suggesting the high anammox performance. Zn-Fe precipitation was observed on the surface of granules during the operation by SEM-EDS, which would impose inhibition to the anammox activity of the granules. Accordingly, it is suggested to relatively reduce the trace metals concentrations, of Fe and Zn in the conventional medium. The findings of this study are expected to be used for a shorter start-up and more stable operation of anammox system. PMID:24455691

  16. Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Palma, Daniel A.P. [CEFET QUIMICA de Nilopolis/RJ, 21941-914 Rio de Janeiro (Brazil)], E-mail: agoncalves@con.ufrj.br; Martinez, Aquilino S.; Goncalves, Alessandro C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)

    2009-09-15

    The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.

  17. Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

    International Nuclear Information System (INIS)

    Palma, Daniel A.P.; Martinez, Aquilino S.; Goncalves, Alessandro C.

    2009-01-01

    The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.

  18. Experience in start-up of the South-Ukrainian-2 power unit with the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Belov, Yu.V.; Kazakov, V.A.; Kirilov, V.V.; Kryakvin, L.V.

    1987-01-01

    The volume, sequence and dates of works fulfilled during hot testing, physical and power start-ups and while bringing output of the South-Ukranian-2 power unit with the WWER-1000 reactor to the design figure are described. The works were fulfilled according to the standard schedules from October in 1984 till April in 1985. Combination of the stages and intensification of works before the physical start-up have allowed to shorten the dates by 90 days as compared to the schedule. The physical and power start-ups, including bringing reactor output to the design figure, were performed during 140 days, that permits to shorten the dates by 20 days more. The results of physical experiments carried out at the South-Ukranian-2 power unit, are in good agreement with the data obtained at the first power units of the given and Kalinin NPPs. Besides, during physical and power start-ups additional measures ensuring nuclear safety are developed

  19. Fundamental study on thermo-hydraulics during start-up in natural circulation boiling water reactors, (1)

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Chiang Jing-Hsien; Takahashi, Tohru; Wataru, Masumi; Mori, Michitsugu.

    1992-01-01

    Recently, many concepts, in which passive and simplified functions are actively adapted, have been proposed for the next generation LWRs. The natural circulation BWR is one such considered from the requirements for next generation LWRs as compared with current BWRs. It is pointed out from this consideration that a thermo-hydraulic instability, which may appear during start-up, greatly influences concept feasibility because its occurence makes operation for raising power output difficult. Thermo-hydraulic instabilities are investigated experimentally under conditions simulating normal and abnormal start-up processes. It is clarified that three kinds of thermo-hydraulic instabilities may occur during start-up in the natural circulation BWR according to its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation instability induced by hydrostatic head fluctuation in steam separators and (3) density wave instability. Driving mechanisms of the geysering and the natural circulation instability, which have never understood enough, are inferred from the results. Finally, the difference of thermo-hydraulic behavior during start-up processes between thermal natural circulation boilers and the Dodewaard reactor is discussed. (author)

  20. Analysis of the IEA-R1 reactor start-up procedures - an application of the HazOp method

    International Nuclear Information System (INIS)

    Sauer, Maria Eugenia Lago Jacques

    2000-01-01

    An analysis of technological catastrophic events that took place in this century shows that human failure and vulnerability of risk management programs are the main causes for the occurrence of accidents. As an example, plants and complex systems where the interface man-machine is close, the frequency of failures tends to be higher. Thus, a comprehensive knowledge of how a specific process can be potentially hazardous is a sine qua non condition to the operators training, as well as to define and implement more efficient plans for loss prevention and risk management. A study of the IEA-R1 research reactor start-up procedures was carried out, based upon the methodology Hazard and Operability Study (HazOp). The analytical and qualitative multidisciplinary HazOp approach provided means to a comprehensive review of the reactor start-up procedures, contributing to improve the understanding of the potential hazards associated to deviations on performing this routine. The present work includes a historical summary and a detailed description of the HazOp technique, as well as case studies in the process industries and the use of expert systems in the application of the method. An analysis of 53 activities of the IEA-R1 reactor start-up procedures was made, resulting in 25 recommendations of changes covering aspects of the project, operation and safety of the reactor. Eleven recommendations have been implemented. (author)

  1. Morphological study of biomass during the start-up period of a fixed-bed anaerobic reactor treating domestic sewage

    OpenAIRE

    Lima,Cláudio Antonio Andrade; Ribeiro,Rogers; Foresti,Eugenio; Zaiat,Marcelo

    2005-01-01

    This work focused on a morphological study of the microorganisms attached to polyurethane foam matrices in a horizontal-flow anaerobic immobilized biomass (HAIB) reactor treating domestic sewage. The experiments consisted of monitoring the biomass colonization process of foam matrices in terms of the amount of retained biomass and the morphological characteristics of the cells attached to the support during the start-up period. Non-fluorescent rods and cocci were found to predominate in the p...

  2. Acclimatization of anaerobic sludge for UASB-reactor start-up

    NARCIS (Netherlands)

    Zeeuw, de W.J.

    1984-01-01

    The Upflow Anaerobic Sludge Bed (UASB) reactor represents a high rate anaerobic wastewater treatment system. The majority of the active biomass in the reactor is present in the form of sludge granules which possess excellent settling properties.
    If no acclimatized (granular)

  3. Preparation of mandatory documentation before the start up of the RA-0 'zero power' nuclear reactor at Cordoba National University

    International Nuclear Information System (INIS)

    Martin, H.R.; Keil, W.M.; Pezzi, N.

    1991-01-01

    Before the start up of the RA-0 'zero power' nuclear reactor installed at Cordoba National University, it was necessary to send to the Regulatory Authority the mandatory documentation which is required in the licensing process. With the previous papers existing for the operation in the first years of the '70, a work program for the future operational training personnel was elaborated. Based on the Authority's applicable rules and the recommendations and with particular criteria originated in the working university conditions, the SAFETY report of RA-0 nuclear reactor was prepared. This paper describes the principal contents, items and documents involved in the safety report. (Author) [es

  4. The use of neutron sources in nuclear reactors start-up after long shutdown periods

    International Nuclear Information System (INIS)

    Ponzoni Filho, P.; Borges, J.B.

    1990-01-01

    The reasons for the use of neutron sources in nuclear reactors, the different kinds of sources used and the alternatives to obtain the required minimum neutron counts in the external source range detectors after long maintenance and refueling periods are presented and discussed. The paper presents a formulation based in physics principles and experimental data, to calculate the power and time of reactor operation required to increase the effective fluence of secondary neutron sources. The option of using actinides produced during operation of the reactor as an additional source of neutrons is also discussed in depth to allow similar calculations in other kinds of reactors. The re-utilization of primary sources is considered as a last option. (author)

  5. Boron mixing transients in a 900 MW PWR vessel for a reactor start-up operation

    International Nuclear Information System (INIS)

    Alvarez, D.; Martin, A.; Schneider, J.P.

    1995-01-01

    In 1991 a R and D action, based on numerical simulations and experiments on PWRs'S primary coolant temperature or boron mixing capabilities, was initiated. This paper presents the test facility BORA-BORA (a 1/5th scaled mock-up of a 900 MW PWR vessel) and the Thermalhydraulic Finite Element Code N3S used for 3D calculations performed on the accurate geometry of the plant. As a validation test case of these experimental and numerical tools, we present the results obtained on the primary coolant mixing capabilities in the vessel with the three loops balanced in mass flow rate. The second part of this report deals with the mixing of a clear water plug in the vessel when a primary coolant pump start-up. The results are obtained in the mock-up in terms of boron concentration at the core inlet for several clear water plug volumes. The numerical results give the complete fluid flow and boron concentration patterns but comparisons were made at the core inlet. (author). 15 refs., 9 figs., 1 tab

  6. Calculation analysis of the neutronic experimental data coming from the NUR reactor start-up

    International Nuclear Information System (INIS)

    Madariaga, M.; Villarino, E.; Relloso, J.; Rubio, R

    1991-01-01

    NUR is a new MTR reactor located in Argelia which became critical in march 1989. It is loaded with a 19 plates LEU Fe. This paper contains: a) Reactivity measurements in the first cores technical information about the Fe and some other data necessary for performing cell and reactor calculations b) calculation comparisons with the measured values (2-D and 3-D calculations) with an statistical analysis of the data set from the control rod calibration. (orig.)

  7. Physicochemical characteristics and microbial community evolution of biofilms during the start-up period in a moving bed biofilm reactor.

    Science.gov (United States)

    Zhu, Yan; Zhang, Yan; Ren, Hong-Qiang; Geng, Jin-Ju; Xu, Ke; Huang, Hui; Ding, Li-Li

    2015-03-01

    This study aimed to investigate biofilm properties evolution coupled with different ages during the start-up period in a moving bed biofilm reactor system. Physicochemical characteristics including adhesion force, extracellular polymeric substances (EPS), morphology as well as volatile solid and microbial community were studied. Results showed that the formation and development of biofilms exhibited four stages, including (I) initial attachment and young biofilm formation, (II) biofilms accumulation, (III) biofilm sloughing and updating, and (IV) biofilm maturation. During the whole start-up period, adhesion force was positively and significantly correlated with the contents of EPS, especially the content of polysaccharide. In addition, increased adhesion force and EPS were beneficial for biofilm retention. Gram-negative bacteria mainly including Sphaerotilus, Zoogloea and Haliscomenobacter were predominant in the initial stage. Actinobacteria was beneficial to resist sloughing. Furthermore, filamentous bacteria were dominant in maturation biofilm. Copyright © 2015 Elsevier Ltd. All rights reserved.

  8. The investigation of fast reactor fuel pin start up behaviour in the irradiation experiment DUELL II

    International Nuclear Information System (INIS)

    Freund, D.; Geithoff, D.

    1988-04-01

    The irradiation experiments DUELL-II within the SNR-300 operational Transient Experimental Program deal with the investigation of fresh mixed oxide fuel behaviour at start-up. The irradiation has been carried out in the HFR Petten in four so-called DUELL capsules with two fuel pin samples each. The fuel pins with a total length of 453 mm contained a fuel column of 150 mm length, consisting of high dense (U,Pu)O 2-x fuel with an initial porosity of 4%, a Pu-content of 20.9%, and an O/Me ratio of 1.96. The fuel pellet diameter was 6.37 mm, the outer diameter of the SS cladding, material No. 1.4970, was 7.6 mm. The irradiation included four phases, consisting of preconditioning at 85% nominal power (corresponds to 550 W/cm), a following increase to full power, and two following full power periods of 1 and 10 days, respectively. Post irradiation examination showed incomplete fuel restructuring in the first capsules with central void diameters of 800 μm in the hot plane, complete restructuring in the last capsule, leading to central voids of approximately 1 mm diameter. The residual gaps between fuel and clad varied between 25 and 44 μm. The clad inner surface did not show any corrosion attack. The analysis of fuel restructuring has been carried out with the computer code SATURN-S showing good agreement with the PIE results. The analysis led to a series of model improvements, especially for crack volume and relocation modelling. (orig./GL) [de

  9. Reactor operating procedures for start up of continuously operated chemical plants

    NARCIS (Netherlands)

    Verwijs, J.W.; Verwijs, J.W.; Kösters, P.H.; van den Berg, Henderikus; Westerterp, K.R.; Kosters, P.G.H.

    1995-01-01

    Rules are presented for the startup of an adiabatic tubular reactor, based on a qualitative analysis of the dynamic behavior of continuously-operated vapor- and liquid-phase processes. The relationships between the process dynamics, operating criteria, and operating constraints are investigated,

  10. Conversion and start up of Tehran Research Reactor with LEU fuel

    International Nuclear Information System (INIS)

    Zaker, M.

    2004-01-01

    The MW Tehran Research Reactor, Highly Enriched Uranium (HEU) fuel has been converted to Low Enriched Uranium (LEU) fuel using U 3 0 8 -Al with less than 20% enriched uranium. Measured value of excess reactivity, control rod worth and other parameters indicate good agreement with computational predictions. (author)

  11. Start-up procedure

    International Nuclear Information System (INIS)

    Marchl, A.; Krebs, W.D.; Aleite, W.

    1975-01-01

    The start-up procedure will be shown on a pressurized water reactor, although most of the activities will occur similarly in other reactor types. The commissioning time can be divided into 5 sections, the phases A to E together lasting 26 months. Subsequently there are a test run of one month and the handling-over of the plant to the operator. A survey of the commissioning sections with several important main events is shown. (orig./TK) [de

  12. Start-up and performance characteristics of a trickle bed reactor degrading toluene

    Directory of Open Access Journals (Sweden)

    Ondrej Misiaczek

    2007-09-01

    Full Text Available The objective of this work was to evaluate toluene degradation in a trickle bed reactor when the loading was carried out by changing the air flow rate. The biofiltration system was inoculated with a mixed microbial population, adapted to degradation of hydrophobic compounds. Polypropylene high flow rings were used as a packing material. The system was operated for a period of 50 days at empty bed residence times ranging from 106s to 13s and with a constant inlet concentration of toluene of 100 mg.m-3. The reactor showed high removal efficiency at higher contact times and increasing elimination capacity with higher air-flow rates. The highest EC value reached was 9.8 gC.m-3.h-1 at EBRT = 13s. During the experiment, the consumption of NaOH solution was also measured. No significant variation of this value was found and an average value of 3.84 mmol of NaOH per gram of consumed carbon was recorded.

  13. Time dependent start-up thermal analysis of a Super Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sutanto,, E-mail: sutanto@fuji.waseda.jp; Oka, Yoshiaki

    2013-10-15

    Highlights: • Time dependent startup thermal analysis of a Super Fast Reactor is performed. • A recirculation system is used for pressurization and for generating supercritical steam. • MCST satisfies the criterion both during subcritical pressure and during power-raising. • MCST is not sensitive to the change of inlet temperature, gap volume and flow rate because of high flow to power ratio. • CHF is not limiting the MCST during subcritical pressure due to large margin of heat flux. -- Abstract: The startup system of a supercritical pressure light water cooled fast reactor (Super FR) is studied by time dependent thermal-hydraulic analysis. The plant analysis code is developed based on an innovative upward flow pattern in all the assemblies of the Super FR. A recirculation system consisting of a steam drum, a circulation pump, and a heat exchanger is used for the startup. Detailed procedures are performed and the maximum cladding surface temperature (MCST) at rated power, 640 °C, is used as the criterion. Firstly a small constant nuclear power is used for rising the core feed water temperature to be 280 °C through the recirculation system. Secondly, pressurization is done in the recirculation system from atmospheric to operating pressure, 25 MPa, by raising the power. Thirdly, line-switching from recirculation mode to once-through direct-cycle is performed while turbines are started by supercritical steam at supercritical pressure. Finally the power is raised to be 100% of power followed by raising the flow rate. During pressurization the heat flux margin is large due to low power used for pressurization and the MCST is much lower than the criterion. The MCST is not sensitive to the inlet temperature, the flow rate, and the gap volume of the core because of high flow to power ratio. Smaller dimension of steam drum can be used for pressurization stably. The MCST satisfies the criterion both during subcritical pressure and during power-raising.

  14. Time dependent start-up thermal analysis of a Super Fast Reactor

    International Nuclear Information System (INIS)

    Sutanto,; Oka, Yoshiaki

    2013-01-01

    Highlights: • Time dependent startup thermal analysis of a Super Fast Reactor is performed. • A recirculation system is used for pressurization and for generating supercritical steam. • MCST satisfies the criterion both during subcritical pressure and during power-raising. • MCST is not sensitive to the change of inlet temperature, gap volume and flow rate because of high flow to power ratio. • CHF is not limiting the MCST during subcritical pressure due to large margin of heat flux. -- Abstract: The startup system of a supercritical pressure light water cooled fast reactor (Super FR) is studied by time dependent thermal-hydraulic analysis. The plant analysis code is developed based on an innovative upward flow pattern in all the assemblies of the Super FR. A recirculation system consisting of a steam drum, a circulation pump, and a heat exchanger is used for the startup. Detailed procedures are performed and the maximum cladding surface temperature (MCST) at rated power, 640 °C, is used as the criterion. Firstly a small constant nuclear power is used for rising the core feed water temperature to be 280 °C through the recirculation system. Secondly, pressurization is done in the recirculation system from atmospheric to operating pressure, 25 MPa, by raising the power. Thirdly, line-switching from recirculation mode to once-through direct-cycle is performed while turbines are started by supercritical steam at supercritical pressure. Finally the power is raised to be 100% of power followed by raising the flow rate. During pressurization the heat flux margin is large due to low power used for pressurization and the MCST is much lower than the criterion. The MCST is not sensitive to the inlet temperature, the flow rate, and the gap volume of the core because of high flow to power ratio. Smaller dimension of steam drum can be used for pressurization stably. The MCST satisfies the criterion both during subcritical pressure and during power-raising

  15. Start up study of UASB reactor treating press mud for biohydrogen production

    International Nuclear Information System (INIS)

    Radjaram, B.; Saravanane, R.

    2011-01-01

    Anaerobic digestion of press mud mixed with water for biohydrogen production was performed in continuous fed UASB bioreactor for 120 days. Experiment was conducted by maintaining constant HRT of 30 h and the volume of biohydrogen evolved daily was monitored. Various parameters like COD, VFA, Alkalinity, EC, Volatile solids, pH with respect to biohydrogen production were monitored at regular interval of time. SBPR was 10.98 ml g -1 COD reduced d -1 and 12.77 ml g -1 VS reduced d -1 on peak yield of biohydrogen. COD reduction was above 70 ± 7%. Maximum gas yield was on the 78th day to 2240 ml d -1 . The aim of the experiment is to study the startup process of UASB reactor for biohydrogen production by anaerobic fermentation of press mud. The inoculum for the process is cow dung and water digested in anaerobic condition for 30 days with municipal sewage sludge. The study explores the viability of biohydrogen production from press mud which is a renewable form of energy to supplement the global energy crisis. -- Highlights: → Feasibility of biohydrogen production from press mud was explored in this study. The gas yield was maximum on the 78th day to 2240 ml d -1 with H 2 % of 52-59%. Biohydrogen yield was about 890 ml kg -1 press mud added d -1 . Press mud is identified as an excellent potential waste to tap energy.

  16. Population dynamics of biofilm development during start-up of a butyrate-degrading fluidized-bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zellner, G.; Geveke, M.; Diekmann, H. (Hannover Univ. (Germany). Inst. fuer Mikrobiologie); Conway de Macario, E. (New York State Dept. of Health, Albany, NY (United States). Wadsworth Center for Laboratories and Research)

    1991-12-01

    Population dynamics during start-up of a fluidized-bed reactor with butyrate or butyrate plus acetate as sole substrates as well as biofilm development on the sand substratum were studied microbiologically, immunologically and by scanning electron microscopy. An adapted syntrophic consortium consisting of Syntrophospora sp., Methanothrix soehngenii, Methanosarcina mazei and Methanobrevibacter arboriphilus or Methanogenium sp. achieved high-rate butyrate degradation to methane and carbon dioxide. Desulfovibrio sp., Methanocorpusculum sp., and Methanobacterium sp. were also present in lower numbers. Immunological analysis demonstrated methanogens antigenically related to Methanobrevibacter ruminantium M1, Methanosarcina mazei S6, M. thermophila TM1, Methanobrevibacter arboriphilus AZ and Methanothrix soehngenii Opfikon in the biofilm. Immunological analysis also showed that the organisms isolated from the butyrate-degrading culture used as a source of inoculum were related to M. soehngenii Opfikon, Methanobacterium formicium MF and Methanospirillum hungatei JF1. (orig.).

  17. Reorganization of the radiologic protection of the nuclear reactor RA-0 for the next starting up at Cordoba National University

    International Nuclear Information System (INIS)

    Martin, H.R.; Chautemps, N.A.; Rumis, D.A.

    1991-01-01

    Due to the fulfillment to the tasks for the new starting up of the RA-0 Nuclear Reactor situated at the National University of Cordoba, it was necessary to plan and organize the service of Radiologic Protection to meet the future requirements in normal operation. The special characteristics that an installation of this type has in the university field, required special attention for making the university staff become aware in the working proceedings to follow up in normal conditions, such as the case of emergency that would originate in the installation. The training of the teaching and non teaching staff of the National University of Cordoba, the adjusting of the installations, the obtention of dosimetry and measurement equipment and the implementation of a monitor system of the staff were the main tasks confronted for the reorganization of the sector. (Author) [es

  18. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

    International Nuclear Information System (INIS)

    Markovic, H.

    1962-07-01

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included [sr

  19. Accelerated formation of hydrogen-producing granules for the start-up of UASB reactors using vinasses

    Directory of Open Access Journals (Sweden)

    César González-Ugalde

    2014-02-01

    Full Text Available Hydrogen-producing granules formation was studied in a CSTR. The aim of this process is to later transfer the mixed liquor to a UASB reactor to reduce its start-up period. Vinasses from a national bioetha­nol-producing industry (from sugar cane were used as substrate and their anaerobic fermentation was carried out under mesophilic conditions. The seed sludge was collected from an UASB reactor oper­ated in an industrial wastewater treatment plant and it was heat treated to inactivate methanogenic bacteria. Total viable and non-viable material growth curves were generated and it was determined that the exponential growth phase of the thermally pre­treated mixed culture was between 20 and 120 h. Finally, the anaerobic fermentation of the vinasses in batch mode for 70 hours, and then in continuous CSTR mode for 7 days, showed to be an effective method for accelerating the formation of hydrogen-producing granules. Using this method, granules with an average size of 1.24 mm were achieved. The good efficiency of the process is attributed to high mass transfer in the CSTR reactor.

  20. Prediction of the stability of BWR reactors during the start-up process; Prediccion de la estabilidad de reactores BWR durante el proceso de arranque

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz E, J.A.; Castillo D, R. [ININ, Km. 36.5 Carretera Mexico-Toluca, 52045 Salazar, Estado de Mexico (Mexico); Blazquez M, J.B. [Centro de Investigaciones Energetics, Medioambientales y Tecnologicas, Av Complutense 22, 28040 Madrid (Spain)

    2004-07-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  1. Morphological study of biomass during the start-up period of a fixed-bed anaerobic reactor treating domestic sewage

    Directory of Open Access Journals (Sweden)

    Cláudio Antonio Andrade Lima

    2005-09-01

    Full Text Available This work focused on a morphological study of the microorganisms attached to polyurethane foam matrices in a horizontal-flow anaerobic immobilized biomass (HAIB reactor treating domestic sewage. The experiments consisted of monitoring the biomass colonization process of foam matrices in terms of the amount of retained biomass and the morphological characteristics of the cells attached to the support during the start-up period. Non-fluorescent rods and cocci were found to predominate in the process of attachment to the polyurethane foam surface. From the 10th week of operation onwards, an increase was observed in the morphological diversity, mainly due to rods, cocci, and Methanosaeta-like archaeal cells. Hydrodynamic problems, such as bed clogging and channeling occurred in the fixed-bed reactor, mainly due to the production of extracellular polymeric substances and their accumulation in the interstices of the bed causing a gradual deterioration of its performance, which eventually led to the system's collapse. These results demonstrated the importance and usefulness of monitoring the dynamics of the formation of biofilm during the start-up period of HAIB reactors, since it allowed the identification of operational problems.Este trabalho apresenta um estudo morfológico de microrganismos aderidos à espuma de poliuretano em reator anaeróbio horizontal de leito fixo (RAHLF, aplicado ao tratamento de esgoto sanitário. O processo de colonização do suporte pela biomassa anaeróbia e as características morfológicas das células aderidas foram monitorados durante o período de partida do reator. Bacilos e cocos não fluorescentes foram predominantes no processo de aderência direta à espuma de poliuretano. Aumento na diversidade biológica foi observado a partir da 10ª semana de operação do reator, com predominância de bacilos, cocos e arqueas metanogênicas semelhantes a Methanosaeta. Problemas hidrodinâmicos, tais como formação de

  2. Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

    Directory of Open Access Journals (Sweden)

    Baschwitz Anne

    2016-01-01

    Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

  3. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  4. An operational protocol for facilitating start-up of single-stage autotrophic nitrogen-removing reactors based on process stoichiometry

    DEFF Research Database (Denmark)

    Mutlu, Ayten Gizem; Vangsgaard, Anna Katrine; Sin, Gürkan

    2013-01-01

    Start-up and operation of single-stage nitritation–anammox sequencing batch reactors (SBRs) for completely autotrophic nitrogen removal can be challenging and far from trivial. In this study, a step-wise procedure is developed based on stoichiometric analysis of the process performance from...

  5. Insertion of control systems models in the Almod 3 computer code for the simulation of Angra I reactor start-up tests

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1981-09-01

    The Almod 3 computer code was modified, aiming at the simulation of Angra I nuclear power plant behavior during some reactor start-up tests. The results obtained with the modified computer code (Almod 3W) are compared with those obtained with the Retran computer code. (E.G.) [pt

  6. Substantiation of operation limits of reactivity insertion during WWER-1000 reactors start-up; Obosnovanie ehkspluatatsionnykh predelov vvoda reaktivnosti pri puske reaktorov WWER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Boev, I; Sabitov, A; Sal` kov, V; Sudarev, O; Yakovlev, A [ATOMTECHENERGO RF, Novovoronezh (Russian Federation)

    1996-12-31

    The methods and programmes used to define the tolerable rate of reactivity insertion during WWER-1000 start-up are presented. They include calculation of the neutron source power in the core during the sub-critical stage and calculation of the relative neutron density and reactor period during the critical stage. The need for optimisation and regulation of tolerable rates is discussed along with the tool parameters affecting the reactivity during start-up. The possibility of increasing the feed rate of pure condensate into the first loop during the time needed to reach critical stage is justified. 4 refs., 3 tabs.

  7. Long life boron-lined proportional counters fitted to the start-up range of the nuclear reactors

    International Nuclear Information System (INIS)

    Dauphin, G.; Duchene, J.

    1983-03-01

    CEA has been engaged in a research and development program in technical cooperation with RTC, to improve the life time of the boron lined proportional counters used for PWR's start up. The present report is a synthesis of these works. The processes by which the discrimination curves show a shift in count rate versus the neutron fluences are described and, the possible ways to get improved performances and the tests performed on the different prototypes are presented. Analysis of the results is given [fr

  8. Start-up of Rapsodie

    International Nuclear Information System (INIS)

    Pontier, R.

    1967-01-01

    After giving a general description of Rapsodie this report presents the conditions in which the start-up occurred and in which the tests were carried out. A chronological account is given of the operations and of the main events which occurred. The modifications made to the reactor during this period are described and a synthesis of the results obtained is presented. (author) [fr

  9. Complementary role of critical integral experiment and power reactor start-up experiments for LMFBR neutronics data and method validation

    International Nuclear Information System (INIS)

    Salvatores, M.

    1986-09-01

    Both critical experiments and power reactor results play at present a complementary role in reducing the uncertainties in Key design parameters for LMFBR, which can be relevant for the economic performances of this type of reactors

  10. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  11. Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs

  12. Fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-01-01

    This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

  13. Lean start-up

    DEFF Research Database (Denmark)

    Rasmussen, Erik Stavnsager; Tanev, Stoyan

    2016-01-01

    The risk of launching new products and starting new firms is known to be extremely high. The Lean Start-up approach is a way of reducing these risks and enhancing the chances for success by validating the products and services in the market with customers before launching it in full scale. The ma...... and the final business model. In other words: The start-up must first nail the problem together with the customers, then develop the solution and test, and then in the end scale it to a full-grown business model.......The risk of launching new products and starting new firms is known to be extremely high. The Lean Start-up approach is a way of reducing these risks and enhancing the chances for success by validating the products and services in the market with customers before launching it in full scale. The main...

  14. FPGA based computation of average neutron flux and e-folding period for start-up range of reactors

    International Nuclear Information System (INIS)

    Ram, Rajit; Borkar, S.P.; Dixit, M.Y.; Das, Debashis

    2013-01-01

    Pulse processing instrumentation channels used for reactor applications, play a vital role to ensure nuclear safety in startup range of reactor operation and also during fuel loading and first approach to criticality. These channels are intended for continuous run time computation of equivalent reactor core neutron flux and e-folding period. This paper focuses only the computational part of these instrumentation channels which is implemented in single FPGA using 32-bit floating point arithmetic engine. The computations of average count rate, log of average count rate, log rate and reactor period are done in VHDL using digital circuit realization approach. The computation of average count rate is done using fully adaptive window size moving average method, while Taylor series expansion for logarithms is implemented in FPGA to compute log of count rate, log rate and reactor e-folding period. This paper describes the block diagrams of digital logic realization in FPGA and advantage of fully adaptive window size moving average technique over conventional fixed size moving average technique for pulse processing of reactor instrumentations. (author)

  15. Starting up the upstarts.

    Science.gov (United States)

    Greene, J

    1997-12-20

    Venture capitalists pour $1 billion a year into health care--and that investment may be the most overlooked indicator of new business opportunities. Signs show that companies focused on consolidation and cost-cutting are off the A list for risk capital. Instead, venture capitalists are targeting start-ups that save money on the front lines by truly managing care.

  16. Experiences from start-up and use of fail-safe reactor level meter with KNITU probes

    International Nuclear Information System (INIS)

    Badiar, S.; Sipka, J.; Vanco, P.; Slanina, Marek; Liska, L.; Gavora, D.

    2001-01-01

    Within the framework of the reconstruction of both V1 Bohunice units with WWER-440/V-230 reactors, the VUJE-SIEMENS consortium implemented measurement systems for coolant level monitoring in reactor pressure vessels with 1E qualification, resp. with the qualification of category 1 according to NUSS RG1.97. The solution uses KNITU 11 probes made by Russian POZIT company. In operating plants, the installation causes problems in relation to the existing technology and quality assurance system. In the phase of implementation, the most important tasks were to resolve component quality, to commission the system, and to check its performance. (Authors)

  17. Fission reactor container

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1991-01-01

    Cooling water is sent without using dynamic equipments upon loss of coolants accident in a pressure vessel by improving an arrangement of a nuclear reactor pressure vessel. That is, a containing space is formed at the center of a suppression chamber for storing cooling water while being partitioned with each other, in which the pressure vessel is placed. Further, a water reservoir is formed above the pressure vessel. Then a water discharge pipe is connected to the reservoir for submerging the stored water over the pressure vessel upon occurrence of loss of coolants accident. Further, a water injection pipe is disposed between the pressure suppression chamber and the pressure vessel for injecting the cooling water in the pressure suppression chamber to the reactor core of the pressure vessel by the difference of a water head upon loss of coolants accident. With such a constitution, the pressure vessel has high earthquake proofness. Further, upon loss of coolants accident of the pressure vessel, the cooling water in the reservoir is discharged to submerge and cool the pressure vessel efficiently. Further, the reactor core of the pressure vessel can certainly be cooled by the cooling water of the pressure suppression chamber without relying on dynamic equipments. (I.S.)

  18. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Bess, John D.; Fujimoto, Nozomu

    2014-01-01

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  19. Fission reactor control rod

    International Nuclear Information System (INIS)

    Irie, Tomoo.

    1991-01-01

    The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

  20. Status report about the works for the start up of the RA-0 'zero power' nuclear reactor at the Cordoba National University

    International Nuclear Information System (INIS)

    Martin, H.R; Carballido, C.; Oliveras, T.

    1991-01-01

    After two years of works at the Cordoba National University for the new start-up of the RA-0 'zero power' nuclear reactor, the results obtained are herein presented. Starting with practically null infrastructure at the beginning, specially in human resources and instrumentation of the reactor, the objectives can be considered satisfactory. The training in work of the future operational staff, the design and the construction of the instrumentation and the fitting of the installations are the principal items described in this paper. An special attention is devoted to the insertion of this type of installation in the university organization, usually not prepared for the quality and control activities, which is necessarily considered in these type of works. (Author) [es

  1. Physical start up of the Dalat nuclear research reactor with the core configuration exempt from neutron trap; Khoi dong vat ly lo phan ung hat nhan Da Lat voi cau hinh vung hoat khong co bay notron

    Energy Technology Data Exchange (ETDEWEB)

    Hien, Pham Duy; Huy, Ngo Quang; Long, Vu Hai; Mai, Tran Khanh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The nominal power of the reconstructed Dalat reactor is of 500 KW. After a meticulous preparation the Russian and Vietnamese teams have proceeded to the physical reactor start-up in November 1983 with the core configuration exempt from the neutron trap. The reactor has reached the physical criticality at 19h50 on 1 November 1983. The report delineates different steps of the start-up procedure. 2 refs., 3 figs., 7 tabs.

  2. Prokaryotic diversity and dynamics in a full-scale municipal solid waste anaerobic reactor from start-up to steady-state conditions.

    Science.gov (United States)

    Cardinali-Rezende, Juliana; Colturato, Luís F D B; Colturato, Thiago D B; Chartone-Souza, Edmar; Nascimento, Andréa M A; Sanz, José L

    2012-09-01

    The prokaryotic diversity of an anaerobic reactor for the treatment of municipal solid waste was investigated over the course of 2 years with the use of 16S rDNA-targeted molecular approaches. The fermentative Bacteroidetes and Firmicutes predominated, and Proteobacteria, Actinobacteria, Tenericutes and the candidate division WWE1 were also identified. Methane production was dominated by the hydrogenotrophic Methanomicrobiales (Methanoculleus sp.) and their syntrophic association with acetate-utilizing and propionate-oxidizing bacteria. qPCR demonstrated the predominance of the hydrogenotrophic over aceticlastic Methanosarcinaceae (Methanosarcina sp. and Methanimicrococcus sp.), and Methanosaetaceae (Methanosaeta sp.) were measured in low numbers in the reactor. According to the FISH and CARD-FISH analyses, Bacteria and Archaea accounted for 85% and 15% of the cells, respectively. Different cell counts for these domains were obtained by qPCR versus FISH analyses. The use of several molecular tools increases our knowledge of the prokaryotic community dynamics from start-up to steady-state conditions in a full-scale MSW reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

  3. Single-phase and two-phase anaerobic digestion of fruit and vegetable waste: Comparison of start-up, reactor stability and process performance

    International Nuclear Information System (INIS)

    Ganesh, Rangaraj; Torrijos, Michel; Sousbie, Philippe; Lugardon, Aurelien; Steyer, Jean Philippe; Delgenes, Jean Philippe

    2014-01-01

    Highlights: • Single-phase and two-phase systems were compared for fruit and vegetable waste digestion. • Single-phase digestion produced a methane yield of 0.45 m 3 CH 4 /kg VS and 83% VS removal. • Substrate solubilization was high in acidification conditions at 7.0 kg VS/m 3 d and pH 5.5–6.2. • Energy yield was lower by 33% for two-phase system compared to the single-phase system. • Simple and straight-forward operation favored single phase process over two-phase process. - Abstract: Single-phase and two-phase digestion of fruit and vegetable waste were studied to compare reactor start-up, reactor stability and performance (methane yield, volatile solids reduction and energy yield). The single-phase reactor (SPR) was a conventional reactor operated at a low loading rate (maximum of 3.5 kg VS/m 3 d), while the two-phase system consisted of an acidification reactor (TPAR) and a methanogenic reactor (TPMR). The TPAR was inoculated with methanogenic sludge similar to the SPR, but was operated with step-wise increase in the loading rate and with total recirculation of reactor solids to convert it into acidification sludge. Before each feeding, part of the sludge from TPAR was centrifuged, the centrifuge liquid (solubilized products) was fed to the TPMR and centrifuged solids were recycled back to the reactor. Single-phase digestion produced a methane yield of 0.45 m 3 CH 4 /kg VS fed and VS removal of 83%. The TPAR shifted to acidification mode at an OLR of 10.0 kg VS/m 3 d and then achieved stable performance at 7.0 kg VS/m 3 d and pH 5.5–6.2, with very high substrate solubilization rate and a methane yield of 0.30 m 3 CH 4 /kg COD fed. The two-phase process was capable of high VS reduction, but material and energy balance showed that the single-phase process was superior in terms of volumetric methane production and energy yield by 33%. The lower energy yield of the two-phase system was due to the loss of energy during hydrolysis in the TPAR and the

  4. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  5. Natural fission reactors - the Oklo phenomenon

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Overview describes the discovery of the site, location of the reactors and site geology and discusses the permanence of fission products, nuclear reaction control mechanisms and trace concentrations of elements that act as poisons. (Author)

  6. Single-phase and two-phase anaerobic digestion of fruit and vegetable waste: Comparison of start-up, reactor stability and process performance

    Energy Technology Data Exchange (ETDEWEB)

    Ganesh, Rangaraj [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Torrijos, Michel, E-mail: michel.torrijos@supagro.inra.fr [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Sousbie, Philippe [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France); Lugardon, Aurelien [Naskeo Environnment, 52 rue Paul Vaillant Couturier, F-92240 Malakoff (France); Steyer, Jean Philippe; Delgenes, Jean Philippe [INRA, UR50, Laboratoire de Biotechnologie de l’Environnement, Avenue des Etangs, Narbonne F-11100 (France)

    2014-05-01

    Highlights: • Single-phase and two-phase systems were compared for fruit and vegetable waste digestion. • Single-phase digestion produced a methane yield of 0.45 m{sup 3} CH{sub 4}/kg VS and 83% VS removal. • Substrate solubilization was high in acidification conditions at 7.0 kg VS/m{sup 3} d and pH 5.5–6.2. • Energy yield was lower by 33% for two-phase system compared to the single-phase system. • Simple and straight-forward operation favored single phase process over two-phase process. - Abstract: Single-phase and two-phase digestion of fruit and vegetable waste were studied to compare reactor start-up, reactor stability and performance (methane yield, volatile solids reduction and energy yield). The single-phase reactor (SPR) was a conventional reactor operated at a low loading rate (maximum of 3.5 kg VS/m{sup 3} d), while the two-phase system consisted of an acidification reactor (TPAR) and a methanogenic reactor (TPMR). The TPAR was inoculated with methanogenic sludge similar to the SPR, but was operated with step-wise increase in the loading rate and with total recirculation of reactor solids to convert it into acidification sludge. Before each feeding, part of the sludge from TPAR was centrifuged, the centrifuge liquid (solubilized products) was fed to the TPMR and centrifuged solids were recycled back to the reactor. Single-phase digestion produced a methane yield of 0.45 m{sup 3} CH{sub 4}/kg VS fed and VS removal of 83%. The TPAR shifted to acidification mode at an OLR of 10.0 kg VS/m{sup 3} d and then achieved stable performance at 7.0 kg VS/m{sup 3} d and pH 5.5–6.2, with very high substrate solubilization rate and a methane yield of 0.30 m{sup 3} CH{sub 4}/kg COD fed. The two-phase process was capable of high VS reduction, but material and energy balance showed that the single-phase process was superior in terms of volumetric methane production and energy yield by 33%. The lower energy yield of the two-phase system was due to the loss of

  7. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  8. The physics of tokamak start-up

    International Nuclear Information System (INIS)

    Mueller, D.

    2013-01-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current

  9. 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.)

  10. Experience in the Kola NPP start-up works

    International Nuclear Information System (INIS)

    Zakharov, S.I.; Omel'chuk, V.V.; Bodrukhin, Yu.M.

    1984-01-01

    Main stages and peculiar features of maintenance and start-up works at WWER-440 type reactor NPPs described. Remarks revealed during complex equipment testing, physical and power start-up will be useful for arrangement of maintenance and start-up works at newly built NPPs

  11. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

    1994-12-31

    Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

  12. Software for physical start-up console

    International Nuclear Information System (INIS)

    Arbet, L.; Suchy, R.

    1991-01-01

    The physical start-up console comprises an PC AT-based control unit equipped with an 80386 processor, and information input/output units. The basic functions to be fulfilled by the control unit software include data acquisition related to the following parameters: neutron physics properties of the reactor core (neutron fluxes recorded by ionization chambers and reactivity recorded by a digital reactimeter), positions of the reactor core control elements (by the digital position meter) and reactor core control measurements, and technological quantities requisite for evaluating physical start-up tests. The measured and calculated data are shown on the control unit display. The setup of the data acquisition system and of user programs is dealt with, and characteristics of the user processes are briefly described. (Z.S.)

  13. Mirror hybrid (fusion--fission) reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.; Neef, W.S.; Devoto, R.S.; Galloway, T.R.; Fink, J.H.; Schultz, K.R.; Culver, D.; Rao, S.

    1977-10-01

    The reference mirror hybrid reactor design performed by LLL and General Atomic is summarized. The reactor parameters have been chosen to minimize the cost of producing fissile fuel for consumption in fission power reactors. As in the past, we have emphasized the use of existing technology where possible and a minimum extrapolation of technology otherwise. The resulting reactor may thus be viewed as a comparatively near-term goal of the fusion program, and we project improved performance for the hybrid in the future as more advanced technology becomes available

  14. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

  15. Contained fissionly vaporized imploded fission explosive breeder reactor

    International Nuclear Information System (INIS)

    Marwick, E.F.

    1978-01-01

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

  16. Start-up tests of NSRR

    International Nuclear Information System (INIS)

    1976-12-01

    Start up tests of the Nuclear Safety Research Reactor (NSRR) were carried out from June 25 to August 15, 1975. The course of tests is in three stage, i.e. critical approach and zero power, power-up and pulse operation. Performance of the reactor was shown to be in good agreement with the design specifications in both steady-state and pulse operations. Test procedures and the results are presented in four parts: (I) general, (II) zero-power tests, (III) power-up tests, and (IV) pulse operation tests. (auth.)

  17. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.; Masson, M.; Briec, M.

    1986-09-01

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 10 13 Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 10 10 Bq (0.5 Ci) per day per ton of fuel

  18. Start-up performance and granular sludge features of an improved external circulating anaerobic reactor for algae-laden water treatment.

    Science.gov (United States)

    Yu, Yaqin; Lu, Xiwu

    2017-09-01

    The microbial characteristics of granular sludge during the rapid start of an enhanced external circulating anaerobic reactor were studied to improve algae-laden water treatment efficiency. Results showed that algae laden water was effectively removed after about 35 d, and the removal rates of chemical oxygen demand (COD) and algal toxin were around 85% and 92%, respectively. Simultaneously, the gas generation rate was around 380 mL/gCOD. The microbial community structure in the granular sludge of the reactor was complicated, and dominated by coccus and filamentous bacteria. Methanosphaera , Methanolinea , Thermogymnomonas , Methanoregula , Methanomethylovorans , and Methanosaeta were the major microorganisms in the granular sludge. The activities of protease and coenzyme F 420 were high in the granular sludge. The intermittent stirring device and the reverse-flow system were further found to overcome the disadvantage of the floating and crusting of cyanobacteria inside the reactor. Meanwhile, the effect of mass transfer inside the reactor can be accelerated to help give the reactor a rapid start.

  19. Start up testing for the secure automated fabrication line

    International Nuclear Information System (INIS)

    Gerber, E.W.; Benson, E.M.; Dahl, R.E.

    1987-01-01

    The secure automated fabrication (SAF) line is a remotely operated, liquid metal reactor fuel fabrication process being built by Westinghouse Hanford Company for the Department of Energy. All process and control equipment is installed and start up testing has been initiated. Start up testing is comprised of five phases, each incorporating higher degrees of equipment integration, automation, and remote control. Testing methodology for SAF line start up is described in this report

  20. Reactor physics and thermodynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1990-01-01

    Neutron kinetics and thermodynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focussed on the properties of the fuel gas, the stationary temperature distribution, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  1. Utilization of fission reactors for fusion engineering testing

    International Nuclear Information System (INIS)

    Deis, G.A.; Miller, L.G.

    1985-01-01

    Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful

  2. Nuclear materials for fission reactors

    International Nuclear Information System (INIS)

    Matzke, H.; Schumacher, G.

    1992-01-01

    This volume brings together 47 papers from scientists involved in the fabrication of new nuclear fuels, in basic research of nuclear materials, their application and technology as well as in computer codes and modelling of fuel behaviour. The main emphasis is on progress in the development of non -oxide fuels besides reporting advances in the more conventional oxide fuels. The two currently performed large reactor safety programmes CORA and PHEBUS-FP are described in invited lectures. The contributions review basic property measurements, as well as the present state of fuel performance modelling. The performance of today's nuclear fuel, hence UO 2 , at high burnup is also reviewed with particular emphasis on the recently observed phenomenon of grain subdivision in the cold part of the oxide fuel at high burnup, the so-called 'rim' effect. Similar phenomena can be simulated by ion implantation in order to better elucidate the underlying mechanism and reviews on high resolution electron microscopy provide further information. The papers will provide a useful treatise of views, ideas and new results for all those scientists and engineers involved in the specific questions of current nuclear waste management

  3. Evaluation of nuclear parameters measurements realized during the Angra-1 reactor start-up tests and the theoretical previsions of these parameters

    International Nuclear Information System (INIS)

    Fernandes, V.B.; Ponzoni Filho, P.; Perrotta, J.A.; Silva Ipojuca, T. da

    1982-01-01

    A summary of measuring techniques and the results of the follow nuclear parameters, are presented: a) reactivity of various control bank; b) capacity of reactor shutdown; c) critical concentrations of soluble boron for various core configurations; d) reactivity isothermal coefficients; e) power mapping; f) power reactivity coefficients. (E.G.) [pt

  4. Search for other natural fission reactors

    International Nuclear Information System (INIS)

    Apt, K.E.; Balagna, J.P.; Bryant, E.A.; Cowan, G.A.; Daniels, W.R.; Vidale, R.J.

    1977-01-01

    Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m 3 and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater than 0.6 b.a. old and if they contained sufficient water but lacked neutron-absorbing impurities. While uranium deposits of northern Canada and northern Australia have received most attention, ore samples have been obtained from the following worldwide locations: the Shinkolobwe and Katanga regions of Zaire; Southwest Africa; Rio Grande do Norte, Brazil; the Jabiluka, Nabarlek, Koongarra, Ranger, and El Sharana ore bodies of the Northern Territory, Australia; the Beaverlodge, Maurice Bay, Key Lake, Cluff Lake, and Rabbit Lake ore bodies and the Great Bear Lake region, Canada. The ore samples were tested for isotopic variations in uranium, neodymium, samarium, and ruthenium which would indicate natural fission. Isotopic anomalies were not detected. Criticality was not achieved in these deposits because they did not have sufficient 235 U content (a function of age and total uranium content) and/or because they had significant impurities and insufficient moderation. A uranium mill monitoring technique has been considered where the ''yellowcake'' output from appropriate mills would be monitored for isotopic alterations indicative of the exhumation and processing of a natural reactor

  5. Loading and initial start-up testing of the low-enrichment uranium core for the Ohio State University research reactor

    International Nuclear Information System (INIS)

    Talnagi, J.W.

    1989-01-01

    Conversion of the Ohio State University Research Reactor (OSURR) from high-enrichment uranium (HEU) fuel to low-enrichment uranium (LEU) fuel elements was begun in August 1985, with funding provided by the U.S. Department of Energy (DOE) and the university. Conversion of the OSURR from HEU to LEU fuel was successfully completed. The reactor is operational at 10-kW steady-state thermal power. Measurements of selected core parameters have been made and compared with predicted values and previous values for the HEU core. In general, measured results agree well with predicted performance, and minor changes have been detected in certain core parameters as a result of the change to LEU fuel. Future plans include additional core testing and a possible increase in operating power

  6. Fusion-Fission hybrid reactors and nonproliferation

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-09-01

    New options for the development of the nuclear energy economy which might become available by a successful development of fusion-breeders or fusion-fission hybrid power reactors, identified and their nonproliferative attributes are discussed. The more promising proliferation-resistance ettributes identified include: (1) Justification for a significant delay in the initiation of fuel processing, (2) Denaturing the plutonium with 238 Pu before its use in power reactors of any kind, and (3) Making practical the development of denatured uranium fuel cycles and, in particular, denaturing the uranium with 232 U. Fuel resource utilization, time-table and economic considerations associated with the use of fusion-breeders are also discussed. It is concluded that hybrid reactors may enable developing a nuclear energy economy which is more proliferation resistant than possible otherwise, whileat the same time, assuring high utilization of t he uranium and thorium resources in an economically acceptable way. (author)

  7. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  8. Development of a real-time simulation tool towards self-consistent scenario of plasma start-up and sustainment on helical fusion reactor FFHR-d1

    Science.gov (United States)

    Goto, T.; Miyazawa, J.; Sakamoto, R.; Suzuki, Y.; Suzuki, C.; Seki, R.; Satake, S.; Huang, B.; Nunami, M.; Yokoyama, M.; Sagara, A.; the FFHR Design Group

    2017-06-01

    This study closely investigates the plasma operation scenario for the LHD-type helical reactor FFHR-d1 in view of MHD equilibrium/stability, neoclassical transport, alpha energy loss and impurity effect. In 1D calculation code that reproduces the typical pellet discharges in LHD experiments, we identify a self-consistent solution of the plasma operation scenario which achieves steady-state sustainment of the burning plasma with a fusion gain of Q ~ 10 was found within the operation regime that has been already confirmed in LHD experiment. The developed calculation tool enables systematic analysis of the operation regime in real time.

  9. Starting up the Saturne synchrotron

    International Nuclear Information System (INIS)

    Salvat, M.

    1958-02-01

    Illustrated by many drawings and graphs, this report describes and comments all operations and measurements to be performed for starting up the Saturne synchrotron until particle acceleration exclusively. The author reports the study of beam as it goes out of the Van de Graaff: experiment of position and stability of the beam axis, study of beam current and geometric characteristics (calibration of the induction probe), experiment of mass separation and proton percentage, and adjustment of regulation and Van de Graaff fall law. In a second part, he reports the optics alignment and the study of optics property (installation of the different sectors, study of inflector end voltage, and influence of inflector position in the chamber). The third part addresses the examination of phenomena associated with injection: injection method and definition of the initial instant, search for injection optimum conditions, study of particle lifetime and of phenomena on the inner probe. The fourth part proposes theoretical additional elements regarding the movement of particles at the injection in the useful area, and phenomena occurring on targets and on the inner probe

  10. Chemical immobilization of fission products reactive with nuclear reactor components

    International Nuclear Information System (INIS)

    Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

    1975-01-01

    This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

  11. Hybrid fission-fusion nuclear reactors

    International Nuclear Information System (INIS)

    Zucchetti, Massimo

    2011-01-01

    A fusion-fission hybrid could contribute to all components of nuclear power - fuel supply, electricity production, and waste management. The idea of the fusion-fission hybrid is many decades old. Several ideas, both new and revisited, have been investigated by hybrid proponents. These ideas appear to have attractive features, but they require various levels of advances in plasma science and fusion and nuclear technology. As a first step towards the development of hybrid reactors, fusion neutron sources can be considered as an option. Compact high-field tokamaks can be a candidate for being the neutron source in a fission-fusion hybrid, essentially due to their design characteristics, such as compact dimensions, high magnetic field, flexibility of operation. This study presents the development of a tokamak neutron source for a material testing facility using an Ignitor-based concept. The computed values show the potential of this neutron-rich device for fusion materials testing. Some full-power months of operation are sufficient to obtain relevant radiation damage values in terms of dpa. (Author)

  12. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  13. Thermal Energetic Reactor with High Reproduction of Fission Materials

    Directory of Open Access Journals (Sweden)

    Vladimir M. Kotov

    2012-01-01

    On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

  14. Preliminary analysis of start up characteristics on SPWR with NESSY (Nuclear ship Engineering Simulation SYstem)

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Kyouya, Masahiko; Sako, Kiyoshi

    1993-09-01

    NESSY (Nuclear ship Engineering Simulation SYstem) has been developed to design advanced marine reactors. SPWR (System integrated PWR) has been designed by JAERI. It doesn't have control rod, and starts up by dilution of boron. we analyzed start up behavior of SPWR by NESSY, and evaluated the safety characteristics on start up and appropriate range of start up rate. (author)

  15. Start-up Strategy for Continuous Bioreactors

    Directory of Open Access Journals (Sweden)

    A.C. da Costa

    1997-06-01

    Full Text Available Abstract - The start-up of continuous bioreactors is solved as an optimal control problem. The choice of the dilution rate as the control variable reduces the dimension of the system by making the use of the global balance equation unnecessary for the solution of the optimization problem. Therefore, for systems described by four or less mass balance equations, it is always possible to obtain an analytical expression for the singular arc as a function of only the state variables. The steady state conditions are shown to satisfy the singular arc expression and, based on this knowledge, a feeding strategy is proposed which leads the reactor from an initial state to the steady state of maximum productivity

  16. Experience in scientific guidance during start-up and normal operation of NPPs with WWER reactors; Opyt nauchnogo rukovodstva puskom i soprovozhdeniya pri ehkspluatatsii ehnergoblokov AEhS s WWER

    Energy Technology Data Exchange (ETDEWEB)

    Abagyan, A; Kazakov, V; Kryakvin, L [All-Russian Inst. for NPP Operation, Moscow (Russian Federation)

    1996-12-31

    The experience gained in the All-Russian Research Institute of NPP Operation (RU) in improvement of reactor start-up and normal operation of NPPs has been reported. The work was carried out on NPPs in Russia, Ukraine, Bulgaria, Hungary and Czechoslovakia. All scientific programmes applied use the system approach, modern methods of safety assessment and the principle of loss effect optimization. For the Kozloduy NPP the following projects have been implemented: design of accelerated unloading system for WWER-1000; refinement of water feeding unit for the WWER-1000 steam generator; response analysis of the safety anti-breakdown system of the second loop. The importance of technical substantiation in the process of decision making is stressed. 6 refs.

  17. Fission fragment assisted reactor concept for space propulsion: Foil reactor

    International Nuclear Information System (INIS)

    Wright, S.A.

    1991-01-01

    The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures

  18. Matching of dense plasma focus devices with fission reactors

    International Nuclear Information System (INIS)

    Harms, A.A.; Heindler, M.

    1978-01-01

    The potential role of dense plasma focus devices as compact neutron sources for fissile fuel breeding in conjunction with existing fission reactors is considered. It is found that advanced plasma focus devices can be used effectively in conjunction with neutronically efficient fission reactors to constitute ''self-sufficient'' breeders. Correlations among the various parameters such as the power output and conversion ratio of the fission reactor with the neutron yield and capacitor bank energy of the dense plasma focus device are presented and discussed

  19. Construction, start-up and operation of a continuously aerated ...

    African Journals Online (AJOL)

    scale SHARON reactor are discussed, along with the construction of the reactor. Special attention is given to the start-up in view of possible toxic effects of high nitrogen concentrations (up to 4 000 mgN·ℓ-1) on the nitrifier population and ...

  20. Innovative fission reactors for this century

    International Nuclear Information System (INIS)

    Minguez, E.

    2007-01-01

    of the 21st Century both innovative fission reactors and fusion reactors. For 2025, it seems that many countries of EU will have to construct NPPs until 40 GWe: France, UK, Germany, North Europe, Russia, Spain, Rumania and Turkey, between others. The viability of these innovative concepts will be presented in this paper

  1. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    Goulo, V.

    1989-06-01

    The document presents the status of nuclear reaction theory concerning optical model development, level density models and pre-equilibrium and direct processes used in calculation of neutron nuclear data for structural materials of fission and fusion reactors. 6 refs

  2. The LOFA analysis of fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Z.-C.; Xie, H.

    2014-01-01

    The fusion-fission hybrid energy reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc, with the fusion neutron source striking the subcritical blanket. The passive safety system, consisting of passive residual heat removal system, passive safety injection system and automatic depressurization system, was adopted into the fusion-fission hybrid energy reactor in this paper. Modeling and nodalization of primary loop, passive core cooling system and partial secondary loop of the fusion-fission hybrid energy reactor using RELAP5 were conducted and LOFA (Loss of Flow Accident) was analyzed. The results of key transient parameters indicated that the PRHRs could mitigate the accidental consequence of LOFA effectively. It is also concluded that it is feasible to apply the passive safety system concept to fusion-fission hybrid energy reactor. (author)

  3. Nuclear data in the problem of fission reactor decommissioning

    International Nuclear Information System (INIS)

    Manokhin, V.N.; Kulagin, N.T.

    1993-01-01

    This report presents a review of the works published in Russia during last several years and devoted to the problem of nuclear data and calculations of nuclear facilities activation for fission reactor decommissioning. 6 refs

  4. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

    Juhl, N.H.; Marwick, E.F.

    1983-01-01

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

  5. Thermal Energetic Reactor with High Reproduction of Fission Materials

    International Nuclear Information System (INIS)

    Kotov, V.M.

    2012-01-01

    Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by 90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

  6. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  7. Chemistry of fission product iodine under nuclear reactor accident conditions

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

  8. On the safety of conceptual fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Okrent, D.; Badham, V.; Caspi, S.; Chan, C.K.; Ferrell, W.J.; Frederking, T.H.K.; Grzesik, J.; Lee, J.Y.; McKone, T.E.; Pomraning, G.C.; Ullman, A.Z.; Ting, T.D.; Kim, Y.I.

    1979-01-01

    A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundent decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered. (Auth.)

  9. 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.

  10. Fission power: a search for a ''second-generation'' reactor

    International Nuclear Information System (INIS)

    Hovingh, J.

    1985-01-01

    This report touches on the history of US fission reactors and explores the current technical status of such reactors around the world, including experimental reactors. Its purpose is to identify, evaluate, and rank the most promising concepts among existing reactors, proposed but unadopted designs, and what can be described as ''new'' concepts. Also discussed are such related concerns as utility requirements and design considerations. The report concludes with some recommendations for possible future LLNL involvement

  11. Start-up analysis for marketing strategy.

    Science.gov (United States)

    Griffith, M J; Baloff, N

    1984-01-01

    The complex start-up effect on utilization of health care services is too often overlooked or underestimated by marketing planners, leading to a range of negative consequences for both the users of services and the provider organization. Start-up analysis allows accurate estimation of these utilization effects for coordinated strategic planning among marketing finance, and operations.

  12. Economic Effect on the Plutonium Cycle of Employing {sup 235}U in Fast Reactor Start-Up; Incidence Economique du Demarrage des Reacteurs Rapides a l'Aide d'Uranium-235 sur le Cycle du Plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Van Dievoet, J.; Egleme, M.; Hermans, L. [BELGONUCLEAIRE, Bruxelles (Belgium)

    1967-09-15

    Preliminary results are presented of a study carried out under an agreement concluded between Euratom and the Belgian Government to evaluate the advantages of loading fast reactors with {sup 235}U. There are several ways of starting up a fast reactor with {sup 235}U: (1) the reactor can be operated entirely with enriched uranium, the plutonium produced being used to start up and operate other reactors; in this case the uranium is recycled within the reactor and more enriched uranium is added; (2) the plutonium produced can be partly recycled within the reactor together with the uranium; in this case the reactor is transformed gradually into a plutonium reactor. These two procedures can be combined and applied simultaneously in different enrichment zones of the same reactor, enriched uranium being added, for example, to the internal zone and plutonium recycled in the external zone. The method of reprocessing the fuel is also a complicating factor, depending on whether the core and the axial breeding blankets are reprocessed together or separately. Similarly, where a reactor has several enrichment zones, these can likewise be reprocessed either together or separately. The calculations are performed with the help of a code that uses the equivalence coefficients defined by Baker and Ross for the part relating to the characteristics of successive reactors, and the discounted fuel cycle cost method for the economic part. In the first stage of this work a rough analysis was made. The reloading of each zone was assumed to be carried out in a single operation, and the time spent by the fuel elements out of pile was ignored. In a later stage, progressive reloading by batches will be considered, with allowance for fabrication and reprocessing times, etc. The most interesting results relate to variations in fuel composition (plutonium content, isotopic composition) from one cycle to another, variations in the fuel cycle characteristics (doubling time, loading and unloading

  13. Start-up of spherical tokamak without a center solenoid

    International Nuclear Information System (INIS)

    Maekawa, Takashi; Nagata, Masayoshi

    2012-01-01

    For low-aspect tokamak reactors, spherical tokamak reactors, ST-type FESF/CTFs, it is essential to remove or minimize a central solenoid (CS). Even with the minimized CS, non-inductive start up of the plasma current is required. Rapid increase in the spontaneous plasma current at the final stage of current start-up drives ignition. At the initial stage, formation of plasma and magnetic surfaces are required. As non-inductive plasma start-up scenarios, ECH/ECCD, LHCD, HHFW, DC HELICITY injection, plasma merging and NBI have been studied. In the present article, the present status and future prospect of experimental and theoretical works on these subjects. (author)

  14. Partida de um reator anaeróbio horizontal para tratamento de efluentes do processamento dos frutos do cafeeiro Start-up of an anaerobic horizontal-flow reactor for treating wastewater from a coffee fruits processing

    Directory of Open Access Journals (Sweden)

    Alisson C. Borges

    2009-01-01

    Full Text Available O presente estudo teve o objetivo de avaliar a partida e a adaptação de um reator anaeróbio horizontal de leito fixo (RAHLF no tratamento de águas residuárias do processamento primário dos frutos do cafeeiro (ARC. O reator foi construído com tubos de PVC de 0,2 m de diâmetro e 3,2 m de comprimento. O sistema foi preenchido com cubos de espuma de poliuretano para imobilização de biomassa ativa. O reator apresentou volume total de 0,1 m³ e volume útil equivalente a 0,04 m³. Em média, houve remoção de 49% da matéria orgânica, com o reator trabalhando sob carga orgânica volumétrica média de 2,66 kg m-3 d-1, medida como DQO. A suplementação de alcalinidade, somada à inoculação prévia de biomassa, proporcionou partida estável do RAHLF, confirmada pelo consumo de ácidos voláteis e adaptação da microbiota ao resíduo. O sistema apresentou resistência às variações de vazão e de carga orgânica observadas, e os teores de fenol e potássio monitorados não causaram inibição da atividade biológica no RAHLF. O maior controle sobre as variações de carga é fator importante na continuidade dos estudos.This study aimed to evaluate the start-up and the adaptation of an anaerobic horizontal-flow immobilized biomass (HAIB reactor in order to treat wastewater from a primary processing of coffee fruits. The reactor was built with PVC tubes of 0.2 m in diameter and 3.2 m in length. The system was filled with cubes of polyurethane foam for immobilization of active biomass. The reactor presented a total capacity of 0.1 m³ and reaction volume equal to 0.04 m³. 49% of organic matter. Removal efficiency was observed, with medium organic volumetric loads equal to 2.66 kg m-3 d-1 (as chemical oxygen demand. The supplementary addition of alkalinity and the previous biomass inoculation provided a stable start-up of the reactor, as confirmed by the reduction of volatile acids and an adaptation of the present microbiology community

  15. Neutronics issues in fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    Liu Chengan

    1995-01-01

    The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

  16. Brief review of the fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1977-01-01

    Much of the conceptual framework of present day fusion-fission hybrid reactors is found in the original work of the early 1950's. Present day motivations for development are quite different. The role of the hybrid reactor is discussed as well as the current activities in the development program

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

    International Nuclear Information System (INIS)

    Martin Deidier, Loick.

    1979-12-01

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

  18. 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

  19. PREFACE: SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor: Devoted to the 75th anniversary of Yu M Ostanevich's birth

    Science.gov (United States)

    Gordely, Valentin; Kuklin, Alexander; Balasoiu, Maria

    2012-03-01

    The Second International Workshop 'SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor', devoted to the 75th anniversary of the birth of Professor Yu M Ostanevich (1936-1992), an outstanding neutron physicist and the founder of small-angle neutron scattering (field, group, and instrument) at JINR FLNPh, was held on 27-30 May at the Frank Laboratory of Neutron Physics. The first Workshop was held in October 2006. Research groups from different neutron centers, universities and research institutes across Europe presented more than 35 oral and poster presentations describing scientific and methodological results. Most of them were obtained with the help of the YuMO instrument before the IBR-2 shutdown in 2006. For the last four years the IBR-2 reactor has been shut down for refurbishment. At the end of 2010 the physical launch of the IBR-2M reactor was finally realized. Nowadays the small-angle neutron scattering (SANS) technique is applied to a wide range of scientific problems in condensed matter, soft condensed matter, biology and nanotechnology, and despite the fact that there are currently over 30 SANS instruments in operation worldwide at both reactor and spallation sources, the demand for beam-time is considerably higher than the time available. It must be remembered, however, that as the first SANS machine on a steady-state reactor was constructed at the Institute Laue Langevin, Grenoble, the first SANS instrument on a 'white' neutron pulsed beam was accomplished at the Joint Institute for Nuclear Research at the IBR-30 reactor, beamline N5. During the meeting Yu M Ostanevich's determinative and crucial contribution to the construction of spectrometers at the IBR-2 high-pulsed reactor was presented, as well as his contribution to the development of the time-of-flight (TOF) small-angle scattering technique, and a selection of other scientific areas. His leadership and outstanding scientific achievements in applications of the

  20. Reference reactor module for NASA's lunar surface fission power system

    International Nuclear Information System (INIS)

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO 2 -fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  1. Lean Start-up in Established Companies

    DEFF Research Database (Denmark)

    Goduscheit, René Chester

    2018-01-01

    Lean start-up is an emergent perspective on how entrepreneurs can bring new products and services to the market. This approach challenges the dominant role of lengthy business plans, linear product development processes, and seeking complete overview of the potential of the new products....../services before market launch. Instead it suggests that start-ups could benefit from a ‘minimum-viable product’ approach where products and services are launched when they contain critical features. The emphasis in the lean start-up approach is on business models rather than the elaborate business plan...... at the companies (strategy meetings, development workshops etc.). The aim is to shed light on the implications for companies that seek to employ lean start-up. These implications will be aimed at aspects like innovation management, organizational structure, customer relations etc....

  2. On fusion and fission breeder reactors

    International Nuclear Information System (INIS)

    Brandt, B.; Schuurman, W.; Klippel, H.Th.

    1981-02-01

    Fast breeder reactors and fusion reactors are suitable candidates for centralized, long-term energy production, their fuel reserves being practically unlimited. The technology of a durable and economical fusion reactor is still to be developed. Such a development parallel with the fast breeder is valuable by reasons of safety, proliferation, new fuel reserves, and by the very broad potential of the development of the fusion reactor. In order to facilitate a discussion of these aspects, the fusion reactor and the fast breeder reactor were compared in the IIASA-report. Aspects of both reactor systems are compared

  3. 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)

  4. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1989-01-01

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

  5. Fission energy: The integral fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Koch, M.; Kazimi, M.S.

    1991-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

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

  8. Smart or Diverse Start-up Teams?

    DEFF Research Database (Denmark)

    Hoogendoorn, Sander; Parker, Simon C.; Van Praag, Mirjam

    2017-01-01

    This paper explores the relationship between cognitive abilities and team performance in a start-up setting. We argue that performance in this setting hinges on three tasks: opportunity recognition, problem solving, and implementation. We theorize that cognitive ability at the individual level has...... others can be assigned to tasks that impose a greater cognitive load (problem solving or opportunity recognition). We present the results of a field experiment in which 573 students in 49 teams started up and managed real companies. We ensured exogenous variation in—otherwise random—team composition...... by assigning students to teams based on their measured cognitive abilities. Each team performed a variety of tasks, often involving complex decision making. The key result of the experiment is that the performance of start-up teams first increases and then decreases with ability dispersion. Strikingly, average...

  9. The Lean and Global Start-up

    DEFF Research Database (Denmark)

    Tanev, Stoyan; Rasmussen, Erik Stavnsager

    For several decades researchers have studied start-up companies with a focus on international markets, suppliers and networks from their inception and on companies that are establishing new, agile business models. This has resulted in two streams of research: The Born Global and International New...... Ventures research and research with a focus on the Lean Start-up company. It is our intention in this paper to give a short presentation of the two research streams and show how they can be merged into one with a focus on newly established technology oriented firms that are lean and global from...

  10. Automated start-up of EBR-II

    International Nuclear Information System (INIS)

    Kisner, R.A.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) and Argonne National Laboratory (ANL) are undertaking a joint project to develop control philosophies, strategies, and algorithms for computer control of the start-up mode of the Experimental Breeder Reactor II (EBR-II). The major objective of this project is to show that advanced liquid-metal reactor (LMR) plants can be operated from low power to full power using computer control. Development of an automated control system with this objective in view will help resolve specific issues and provide proof through demonstration that automatic control for plant start-up is feasible. This paper describes the approach that will be used to develop such a system and some of the features it is expected to have. Structured, rule-based methods, which will provide start-up capability from a variety of initial plant conditions and degrees of equipment operability, will be used for accomplishing mode changes during plant start-up. Several innovative features will be incorporated such as signal, command, and strategy validation to maximize reliability, flexibility to accommodate a wide range of plant conditions, and overall utility. Continuous control design will utilize figures of merit to evaluate how well the controller meets the mission requirements. The operator interface will have unique ''look ahead'' features to let the operator see what will happen next. 15 refs., 7 figs., 1 tab

  11. High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bell, Zane W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, Dominic R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lance, Michael J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Warmack, Robert J. Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Mark J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

  12. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  13. Updated comparison of economics of fusion reactors with advanced fission reactors

    International Nuclear Information System (INIS)

    Delene, J.G.

    1990-01-01

    The projected cost of electricity (COE) for fusion is compared with that from current and advanced nuclear fission and coal-fired plants. Fusion cost models were adjusted for consistency with advanced fission plants and the calculational methodology and cost factors follow guidelines recommended for cost comparisons of advanced fission reactors. The results show COEs of about 59--74 mills/kWh for the fusion designs considered. In comparison, COEs for future fission reactors are estimated to be in the 43--54 mills/kWh range with coal-fired plant COEs of about 53--69 mills/kWh ($2--3/GJ coal). The principal cost driver for the fusion plants relative to fission plants is the fusion island cost. Although the estimated COEs for fusion are greater than those for fission or coal, the costs are not so high as to preclude fusion's competitiveness as a safe and environmentally sound alternative

  14. Thermochemical data for reactor materials and fission products

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1990-01-01

    This volume presents a collection of critically assessed data on inorganic compounds which are of special interest in nuclear reactor safety studies. Thermodynamic equilibrium calculations are an important and widely used instrument in the understanding of the chemical behavior and release of fission products in the course of nuclear reactor accidents. The reliability of such calculations is, nevertheless, limited by the availability of accurate input data for relevant compounds

  15. Systems study of tokamak fusion--fission reactors

    International Nuclear Information System (INIS)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

  16. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  17. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  18. Safety device for nuclear fission reactors

    International Nuclear Information System (INIS)

    Brownlee, M.L.

    1982-01-01

    A plurality of radially arranged and neutron absorbing baffles are stacked in vertical sets under the fuel core assemblies, and the whole enclosed in a bottle shaped containment vessel. The radially arranged baffles of each set extend vertically, and each set has double the number of baffles as the set above it in the stack. A melt-down of a fuel core assembly drops the fissioning nuclear fuel into the stacked sets of baffles, there, as it passes through, to be progressively divided, redivided and dispersed in smaller and smaller masses between the doubling number of baffles in safe fuel pellet size. Neutron absorbing containment prevents contamination of the environment and together with cooling means stops fissioning of fuel

  19. From Jam to Start-up

    DEFF Research Database (Denmark)

    Bering Kjæhr, Emil; Lyngbye Hvid Jensen, Jane; Schoenau-Fog, Henrik

    2015-01-01

    the foundation for new start-up companies. There are, however, an even larger group of very creative and innovative games and artifacts with great potential that go unpublished. This potentially leads to a loss of entrepreneurial opportunities and ultimately the jobs and careers such games could have fostered...

  20. Starting Up after 50. CELCEE Digest.

    Science.gov (United States)

    Seymour, Nicole

    Researchers are finding that older entrepreneurs are an increasing population in many Western countries. It is important to distinguish between entrepreneurs who have simply reached the age of 50 versus those who start up businesses after this age. The latter group is of particular interest because these people have presumably never faced the…

  1. Neutron and thermal dynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1989-01-01

    In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  2. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

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

    1982-12-01

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

  3. Laser-start-up system for magnetic mirror fusion

    International Nuclear Information System (INIS)

    Frank, A.M.; Thomas, S.R.; Denhoy, B.S.; Chargin, A.K.

    1976-01-01

    A CO 2 laser system has been developed at LLL to provide hot start-up plasmas for magnetic mirror fusion experiments. A frozen ammonia pellet is irradiated with a laser power density in excess of 10 13 W/cm 2 in a 50-ns pulse. This system uses commercially available laser systems. Optical components were fabricated both by direct machining and standard techniques. The technologies used in this system are directly applicable to reactor scale systems

  4. Conceptual Analysis of Fission Fragment Magnetic Collimator Reactors

    International Nuclear Information System (INIS)

    Tsvetkov, Pavel V.; Parish, Theodore A.

    2002-01-01

    As part of the current research work within the US DOE NERI Direct Electricity Conversion (DEC) Project on methods for utilizing direct electricity conversion in nuclear reactors, a detailed study of a Fission Fragment Magnetic Collimator Reactor (FFMCR) has been performed. The FFMCR concept is an advanced DEC system that combines advantageous design solutions proposed for application in both fission and fusion reactors. The present study was focused on determining the electrical efficiency and other important operational aspects of the FFMCR concept. In principle, acceptable characteristics have been demonstrated, and results obtained are presented in the paper. Technological visibility of the FFMCR concept and required further design development are discussed. Preliminary characteristics of the promising design are outlined. (authors)

  5. Developments and Tendencies in Fission Reactor Concepts

    Science.gov (United States)

    Adamov, E. O.; Fuji-Ie, Y.

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

  6. Status on potential of advanced fission reactors

    International Nuclear Information System (INIS)

    L-Zaleski, C.P.

    1978-01-01

    In this short lecture, only two types of reactors will be discussed: the liquid metal fast breeder reactors (LMFBR) and the high temperature reactors (HTR). This does not mean that other very interesting concepts do not exist, but there are or proven light water reactors and heavy water reactors or has not reached the state of industrial development like molten-salt or gas breeder reactors. In discussing any types of industrial development, it seems to me useful, first to indicate the reasons or motivations for this development. Then I will give a short historical review and analysis of what has been done up to now. For HTR's a very brief status report will be presented. For LMFBR's, I will give indications of experience gained with demonstration plants and more specifically with Phenix, before listing the most important technical problems which still need more work to be fully solved. Finally, I will briefly discuss the economic status and perspectives of LMFBR's and will mention the public acceptance problem

  7. Secondary Circuit Start Up Chemistry Optimisation

    International Nuclear Information System (INIS)

    Fontan, Guillaume; Morel, Pascal

    2012-09-01

    In a context of investment and renewal of equipment, Electricite De France (EDF) put enhanced efforts on operating practices during start-up of the secondary circuit, in order to improve operational performance and materials lifetime. This article focuses on the objective of optimizing the filling, the chemical conditioning and the thermal conditioning of the secondary fluid, while taking into account the following issues: - Limiting the time required to obtain a proper chemistry, - Limiting the amount of water and steam used, - Limiting the amount of effluent generated. The scope is all start-up conditions of secondary circuit, both after refuelling outage or fortuitous shutdowns of the plant. The recommendations produced are based on existing local procedures and good practices, which were collected and developed in order to propose a generic methodology understandable and useful both for operators, chemists and managers. (authors)

  8. Lean and global technology start-ups

    DEFF Research Database (Denmark)

    Tanev, Stoyan; Rasmussen, Erik Stavnsager; Zijdemans, Erik

    2015-01-01

    In this paper the authors introduce the concept of Lean Global Start-up (LGS) as a way of emphasizing the impossibility for new technology start-ups to deal separately with business development, innovation and early internationalization. For a newly established technology firm the task of being...... global and innovative at the same time should be seen as one process. The paper has two components – an introductory conceptual part and an empirical part that should be considered as basis for the preliminary validation of the conceptual insights. The research sample includes six firms – three from...... Canada and three from Denmark. The firms have had significant problems with the complexity, uncertainties and risks of being innovative on a global scale. They have however found ways of addressing these problems by a disciplined knowledge sharing and IP protection strategy and the efficient use...

  9. Business Models for Start up Business

    OpenAIRE

    Boban, Nitin

    2010-01-01

    Gamingdom is a new start up venture that provides online gaming service utilising the latest cloud computing technology. The high demand, popularity and exponential growth in the number of gaming enthusiasts and the market have brought about this innovative idea. This new venture aims to provide an easy method of gaming through the internet without heavy expenses on hardware and software and also minimising the existing high rate of piracy. A business model is an essential tool for buildin...

  10. An analysis of the sliding pressure start-up of SCWR

    International Nuclear Information System (INIS)

    Wang, F.; Yang, J.; Li, H.; Zhang, Y.; Zhang, J.; Shan, J.; Gou, J.; Zhang, B.; Chen, C.

    2012-01-01

    In this paper, the preliminary sliding pressure start-up system and scheme of supercritical water-cooled reactor in CGNPC (CGN-SCWR) were proposed. Thermal-hydraulic behavior in start-up procedures was analyzed in detail by employing advanced reactor subchannel analysis software ATHAS. The maximum cladding temperature (MCT for short) and core power of fuel assembly during the whole start-up process were investigated comparatively. The results show that the recommended start-up scheme meets the design requirements from the perspective of thermal-hydraulic. (authors)

  11. Neutron Tests at the Start-Up of EDF1; Les essais neutroniques au demarrage du reacteur EDF1; Nejtronnye izmereniya pri puske reaktora EDF1; Ensayos neutronicos efectuados durante la puesta en marcha del reactor EDF1

    Energy Technology Data Exchange (ETDEWEB)

    Teste du Bailler, A. [Centre d' Etudes Nucleaires de Saclay (France); Janin, R. [Electricite de France, Paris (France)

    1963-10-15

    A series of neutron measurements, for which the principal experimental methods perfected at the Marcoule reactors were used, was carried out at the start-up of EDF1. The measurements were designed mainly to determine the efficiency of the control rods at different depths of insertion. From them a rod-withdrawal configuration was derived which allowed full-power operation without infringing certain limitations on cladding and gas temperatures. At the same time flux measurements were made for different shim-rod positions and different absorber loadings in certain channels. These measurements based on preliminary two-dimensional calculations, were obtained by activation of point detectors,using the standard technique of air poisoning. At certain temperature plateaus (up to 140{sup o}C), measurements of temperature coefficients and control-rod efficiency were made. Spectrum index measurements were carried out at the same time by activation of appropriate detectors (U, Pu, Lu, Mn, In, Au). The oscillation technique was used to measure the efficiency of certain shim rods. Finally, fast-neutron measurements were made in connection with studies of shielding and graphite damage. (author) [French] Une serie de mesures neutroniques utilisant les principales methodes experimentales mises au point sur les reacteurs de Marcoule a ete effectuee au cours du demarrage d'EDF1. Les mesures portent essentiellement sur l 'efficacite des barres de controle a differents enfoncements. On en deduit une configuration de montee des barres permettant d'obtenir la pleine puissance en respectant certaines limitations sur les temperatures de gaines et de gaz. Parallelement des mesures de flux ont ete faites pour differentes positions des barres de compensation et pour divers chargements d'absorbants dans certains canaux, suivant des calculs previsionnels a deux dimensions. Ces mesures sont obtenues par activation de detecteurs ponctuels, au moyen de la technique classique par empoisonnement a l

  12. Fission product release from TRIGA-LEU reactor fuels

    International Nuclear Information System (INIS)

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

    1980-01-01

    Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

  13. Start up testing for the secure automated fabrication line

    International Nuclear Information System (INIS)

    Gerber, E.W.; Benson, E.M.; Dahl, R.E.

    1986-01-01

    The Secure Automated Fabrication (SAF) Line has been designed and built by Westinghouse Hanford Company for the Department of Energy at the Hanford Site near Richland, Washington. The SAF Line will provide the capability for remote manufacture of fuel for Liquid Metal Reactors, and will supply fuel for the Fast Flux Test Facility (FFTF). The SAF process is highly automated and represents a major advancement in nuclear fuel manufacturing, offering significant improvements in product quality, productivity, safety, and accountability of Special Nuclear Materials. The construction phase of the project is complete, and testing has been initiated to accomplish start up of the plant for manufacture of FFTF fuel. This paper describes the test methodology used for SAF Line start up

  14. Fission reactor recycling pump handling device

    International Nuclear Information System (INIS)

    Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru.

    1991-01-01

    This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.)

  15. Fission reactor recycling pump handling device

    Energy Technology Data Exchange (ETDEWEB)

    Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru

    1991-06-24

    This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.).

  16. Participation in testing and start up operations

    International Nuclear Information System (INIS)

    Prasad, Y.S.R.

    1977-01-01

    Testing and start up operations of a nuclear power plant require careful planning. A detailed program of tests and the responsibility of implementing them is discussed. Requirement of documentation covering the tests and operating procedures is explained. The performance of the system during normal and abnomal operating conditions is analysed and required are modifications carried out. Various phases of commissioning and their significance are explained. Preparation of maintenance documentation and training of operating and maintenance staff during this period are discussed. Necessity of close liaison between the regulatory body and the operating organization is explained. (orig.) [de

  17. Overview of standards subcommittee 8, fissionable materials outside reactors

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1996-01-01

    The American Nuclear Society's Standards Subcommittee 8, titled open-quotes Fissionable Materials Outside Reactors,close quotes has worked for the past 35 yr to prepare and promote standards on nuclear criticality safety for the handling, processing, storing, and transportation of fissionable materials outside reactors. The reader is referred to the Transactions of the American Nuclear Society, Vols. 39 (1981) and 64 (1991), for previous papers associated with ANS-8 poster sessions. In addition to discussions on the then-current standards, the reader will find articles on working group efforts that never materialized into standards, such as proposed 8.13, open-quotes Use of the Solid-Angle Method in Nuclear Criticality Safety,close quotes and on applications and critiques of current standards. The paper by McLendon in Vol. 39 is particularly interesting as an overview of the early history of ANS-8 and its standards

  18. Expected value of finite fission chain lengths of pulse reactors

    International Nuclear Information System (INIS)

    Liu Jianjun; Zhou Zhigao; Zhang Ben'ai

    2007-01-01

    The average neutron population necessary for sponsoring a persistent fission chain in a multiplying system, is discussed. In the point reactor model, the probability function θ(n, t 0 , t) of a source neutron at time t 0 leading to n neutrons at time t is dealt with. The non-linear partial differential equation for the probability generating function G(z; t 0 , t) is derived. By solving the equation, we have obtained an approximate analytic solution for a slightly prompt supercritical system. For the pulse reactor Godiva-II, the mean value of finite fission chain lengths is estimated in this work and shows that the estimated value is reasonable for the experimental analysis. (authors)

  19. Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

  20. Role of fission gas release in reactor licensing

    International Nuclear Information System (INIS)

    1975-11-01

    The release of fission gases from oxide pellets to the fuel rod internal voidage (gap) is reviewed with regard to the required safety analysis in reactor licensing. Significant analyzed effects are described, prominent gas release models are reviewed, and various methods used in the licensing process are summarized. The report thus serves as a guide to a large body of literature including company reports and government documents. A discussion of the state of the art of gas release analysis is presented

  1. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  2. Transient fission product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.

    1995-01-01

    Sweep gas experiments performed at CRL from 1979 to 1985 have been analysed to determine the fraction of the fission product gas inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the xenon release from companion fuel elements and from a well documented experimental fuel bundle irradiated in the NRU reactor. The measured gas release could be matched to within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. (author)

  3. Fission-product burnup chain model for research reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

    1990-12-01

    A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

  4. Structural materials issues for the next generation fission reactors

    Science.gov (United States)

    Chant, I.; Murty, K. L.

    2010-09-01

    Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

  5. Fusion--fission hybrid reactors based on the laser solenoid

    International Nuclear Information System (INIS)

    Steinhauer, L.C.; Taussig, R.T.; Quimby, D.C.

    1976-01-01

    Fusion-fission reactors, based on the laser solenoid concept, can be much smaller in scale than their pure fusion counterparts, with moderate first-wall loading and rapid breeding capabilities (1 to 3 tonnes/yr), and can be designed successfully on the basis of classical plasma transport properties and free-streaming end-loss. Preliminary design information is presented for such systems, including the first wall, pulse coil, blanket, superconductors, laser optics, and power supplies, accounting for the desired reactor performance and other physics and engineering constraints. Self-consistent point designs for first and second generation reactors are discussed which illustrate the reactor size, performance, component parameters, and the level of technological development required

  6. Thermodynamic cycle calculations for a pumped gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.

    1991-01-01

    Finite and 'infinitesimal' thermodynamic cycle calculations have been performed for a 'solid piston' model of a pumped Gaseous Core Fission Reactor with dissociating reactor gas, consisting of Uranium, Carbon and Fluorine ('UCF'). In the finite cycle calculations the influence has been investigated of several parameters on the thermodynamics of the system, especially on the attainable direct (nuclear to electrical) energy conversion efficiency. In order to facilitate the investigation of the influence of dissociation, a model gas, 'Modelium', was developed, which approximates, in a simplified, analytical way, the dissociation behaviour of the 'real' reactor gas. Comparison of the finite cycle calculation results with those of a so-called infinitesimal Otto cycle calculation leads to the conclusion that the conversion efficiency of a finite cycle can be predicted, without actually performing the finite cycle calculation, with reasonable accuracy, from the so-called 'infinitesimal efficiency factor', which is determined only by the thermodynamic properties of the reactor gas used. (author)

  7. Fission product poisoning in KS-150 reactor operation

    International Nuclear Information System (INIS)

    Rana, S.B.

    1978-01-01

    A three-dimensional model of the KS-150 reactor was used to study reactivity changes induced by reactor poisoning with fission products Xe 135 and Sm 149 . A comparison of transients caused by the poisoning showed the following differences: (1) the duration of the transient Xe poisoning (2 days) is shorter by one order of magnitude than the duration of Sm poisoning (20 days); however, the level of Xe poisoning is greater approximately by one order than the level of the Sm poisoning; (2) the level of steady-state Xe poisoning depends on the output level of the reactor; steady-state Sm poisoning does not depend on this level; (3) following reactor shutdown Xe poisoning may increase to the maximum value of up to Δrhosub(Xe)=20% and will then gradually decrease; Sm poisoning may reach maximum values of up to Δrhosub(Sm)=2% and does not decrease. (J.B.)

  8. Nuclear data requirements for fission reactor decommissioning

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1993-01-01

    The meeting was attended by 13 participants from 8 Member States and 2 International Organizations who reviewed the status of the nuclear data libraries and computer codes used to calculate the radioactive inventory in the reactor unit components for the decommissioning purposes. Nuclides and nuclear reactions important for determination of the radiation fields during decommissioning and for the final disposal of radioactive waste from the decommissioned units were identified. Accuracy requirements for the relevant nuclear data were considered. The present publication contains the text of the reports by the participants and their recommendations to the Nuclear Data Section of the IAEA. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  9. Action plan for the task: Physical measurements at the RA reactor related to VISA-2 project, '0' program, Reactor start-up and measurement of basic parameters of the new core; Plan rada po zadatku: Fizicka merenja na reaktoru RA u vezi projekta VISA-2, '0' program, Pustanje u rad reaktora RA i merenje osnovnih parametara novog jezgra

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-01

    This report consists of two parts. Part one describes the RA reactor start-up, measurements of thermal neutron flux distribution, measurements of epithermal flux, fast neutron flux distribution, absolute values of both thermal and fast neutron fluxes, calibration of regulating rods, and measurements of neutron flux inside the fuel elements. All the mentioned measurements were done at low power level. Part two includes description of the reactor power increase up to nominal value of 6.5 MW, and measurements of thermal neutron flux distribution under xenon poisoning conditions, measurements of epithermal neutrons, absolute values of both thermal and fast neutron fluxes, and measurements of thermal and epithermal neutron fluxes at the exit of the horizontal experimental channel HK-d.

  10. The design of control algorithm for automatic start-up model of HWRR

    International Nuclear Information System (INIS)

    Guo Wenqi

    1990-01-01

    The design of control algorithm for automatic start-up model of HWRR (Heavy Water Research Reactor), the calculation of μ value and the application of digital compensator are described. Finally The flow diagram of the automatic start-up and digital compensator program for HWRR are given

  11. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters; Fizicka merenja na reaktoru RA u vezi projekta VISA-2 - I deo, Pustanje u rad reaktora RA i merenje fizickih parametara novog jezgra reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included. [Serbo-Croat] Svrha merenja je odredjivanje neutronskog fluksa u reaktoru RA. S obzirom na uvecani broj tehnoloskih kanala of 56 na 68 u vezi projekta VISA-2, bilo je potrebno ponovo dovesti reaktora RA do kriticnosti i izvrsiti merenja karakteristika fluksa neutrona. Posebno je pripremljen 'program pustanja u pogon reaktora RA', koji je sadrzan u ovom dokumentu. Program merenja bio je podeljen na dve faze. Prva faza je merenje fluksa pre podizanju reaktora na nominalnu snagu. Slicna merenja vrsena su i na vecim snagama u drugoj fazi, pod uslovima ravnoteznog zatrovanja reaktora ksenonom, jer se tada pokazuju izvesne promene u odgovarajucim karakteristikama fluksa neutrona. Ovaj izvestaj sadrzi merene vrednosti raspodele fluksa i apsolutne vrednosti termalnih i brzih neutrona kao i kadmijumskih odnosa koji su korisceni za odredjivanje fluksa epitermalnih neutrona. Opisana je kalibracija regulacionih sipki za hladan nezatrovan reaktor.

  12. Start-up fuel and power flattening of sodium-cooled candle core

    International Nuclear Information System (INIS)

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-01-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

  13. Neutron dosimetry for radiation damage in fission and fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1979-01-01

    The properties of materials subjected to the intense neutron radiation fields characteristic of fission power reactors or proposed fusion energy devices is a field of extensive current research. These investigations seek important information relevant to the safety and economics of nuclear energy. In high-level radiation environments, neutron metrology is accomplished predominantly with passive techniques which require detailed knowledge about many nuclear reactions. The quality of neutron dosimetry has increased noticeably during the past decade owing to the availability of new data and evaluations for both integral and differential cross sections, better quantitative understanding of radioactive decay processes, improvements in radiation detection technology, and the development of reliable spectrum unfolding procedures. However, there are problems caused by the persistence of serious integral-differential discrepancies for several important reactions. There is a need to further develop the data base for exothermic and low-threshold reactions needed in thermal and fast-fission dosimetry, and for high-threshold reactions needed in fusion-energy dosimetry. The unsatisfied data requirements for fission reactor dosimetry appear to be relatively modest and well defined, while the needs for fusion are extensive and less well defined because of the immature state of fusion technology. These various data requirements are examined with the goal of providing suggestions for continued dosimetry-related nuclear data research

  14. Hefei experimental hybrid fusion-fission reactor conceptual design

    International Nuclear Information System (INIS)

    Qiu Lijian; Luan Guishi; Xu Qiang

    1992-03-01

    A new concept of hybrid reactor is introduced. It uses JET-like(Joint European Tokamak) device worked at sub-breakeven conditions, as a source of high energy neutrons to induce a blanket fission of depleted uranium. The solid breeding material and helium cooling technique are also used. It can produce 100 kg of 239 Pu per year by partial fission suppressed. The energy self-sustained of the fusion core is not necessary. Plasma temperature is maintained by external 20 MW ICRF (ion cyclotron resonance frequency) and 10 MW ECRF (electron cyclotron resonance frequency) heating. A steady state plasma current at 1.5 Ma is driven by 10 MW LHCD (lower hybrid current driven). Plasma density will be kept by pellet injection. ICRF can produce a high energy tail in ion distribution function and lead to significant enhancement of D-T reaction rate by 2 ∼ 5 times so that the neutron source strength reaches to the level of 1 x 10 19 n/s. This system is a passive system. It's power density is 10 W/cm 3 and the wall loading is 0.6 W/cm 2 that is the lower limitation of fusion and fission technology. From the calculation of neutrons it could always be in sub-critical and has intrinsic safety. The radiation damage and neutron flux distribution on the first wall are also analyzed. According to the conceptual design the application of this type hybrid reactor earlier is feasible

  15. Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap; Khoi dong vat ly lo phan ung hat nhan Da Lat voi cau hinh vung hoat co bay notron

    Energy Technology Data Exchange (ETDEWEB)

    Hien, Pham Duy; Huy, Ngo Quang; Long, Vu Hai; Mai, Tran Khanh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs.

  16. Transient fission-product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

    1997-12-01

    Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

  17. JENDL-4.0 benchmarking for fission reactor applications

    International Nuclear Information System (INIS)

    Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

    2011-01-01

    Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

  18. The electronuclear cycle: from fission to new reactor systems

    International Nuclear Information System (INIS)

    Belier, G.; Cugnon, J.; Lapoux, V.; Liatard, E.; Porquet, Marie-Genevieve; Rudolf, G.

    2006-09-01

    The Joliot Curie School trains each year, and since 1981, PhD students, post-Doctorates and researchers on scientific breakthroughs performed in a topic related to nuclear physics, in a broad range. These proceedings brings together the 11 lectures given at the 2006 session of Joliot Curie School on the topic of the electronuclear cycle: - Fission: from phenomenology to theory (Berger, J.F.); - Physics of nuclear reactors (Baeten, P.); - Data modeling and evaluation (Bauge, E.; Hilaire, S.); - Measurement of cross sections of interest for minor actinides incineration (Jurado, B.); - Spallation data and modelling for hybrid reactors (Boudard, A.); - Nuclear wastes: overview (Billard, I.); - Long living nuclear wastes transmutation processes and feasibility (Varaine, F.); - Hybrid reactors: recent advances for a demonstrator (Billebaud, A.); - Systems of the future and strategy (David, S.); - Non-nuclear energies (Nifenecker, H.); - Fundamental physics with ultracold neutrons (Protasov, K). The last section is a compilation of abstracts of presentations given by Young searchers' (Young searchers' seminars)

  19. Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

    International Nuclear Information System (INIS)

    Bacher, P.

    1997-01-01

    A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

  20. Photofission observations in reactor environments using selected fission-product yields

    International Nuclear Information System (INIS)

    Gold, R.; Ruddy, F.H.; Roberts, J.H.

    1982-01-01

    A new method for the observation of photofission in reactor environments is advanced. It is based on the in-situ observation of fission product yield. In fact, at a given in-situ reactor location, the fission product yield is simply a weighted linear combination of the photofission product yield, Y/sub gamma/, and the neutron induced fission product yield, Y/sub n. The weight factors arising in this linear combination are the photofission fraction and neutron induced fission fraction, respectively. This method can be readily implemented with established techniques for measuring in-situ reactor fission product yield. For example, one can use the method based on simultaneous irradiation of radiometric (RM) and solid state track recorder (SSTR) fission monitors. The sensitivity and accuracy and current knowledge of fission product yields. Unique advantages of this method for reactor applications are emphasized

  1. Bauxite slurry pipeline: start up operation

    Energy Technology Data Exchange (ETDEWEB)

    Othon, Otilio; Babosa, Eder; Edvan, Francisco; Brittes, Geraldo; Melo, Gerson; Janir, Joao; Favacho, Orlando; Leao, Marcos; Farias, Obadias [Vale, Rio de Janeiro, RJ (Brazil); Goncalves, Nilton [Anglo Ferrous Brazil S.A., Rio de Janeiro, RJ (Brazil)

    2009-07-01

    The mine of Miltonia is located in Paragominas-PA, in the north of Brazil. Bauxite slurry pipeline starts at the Mine of Miltonia and finishes in the draining installation of Alunorte refinery at the port of Barcarena-PA, located approximately 244km away from the mine. The pipeline runs over seven cities and passes below four great rivers stream beds. The system was designed for an underground 24 inches OD steel pipe to carry 9.9 million dry metric tonnes per annum (dMTAs) of 50.5% solid concentration bauxite slurry, using only one pumping station. The system is composed by four storage tanks and six piston diaphragm pumps, supplying a flow of 1680 m3/h. There is a cathodic protection system along the pipeline extension to prevent external corrosion and five pressure monitoring stations to control hydraulic conditions, there is also a fiber optic cable interconnection between pump station and terminal station. Pipeline Systems Incorporated (PSI) was the designer and followed the commissioning program of the start up operations. This paper will describe the beginning of the pipeline operations, technical aspects of the project, the operational experiences acquired in these two years, the faced problems and also the future planning. (author)

  2. Giving start-ups a helping hand

    CERN Multimedia

    2008-01-01

    The French-Swiss foundation for research and technology (FFSRT) joined forces with CERN to organise an information day on setting up new businesses, at the Globe of Science and Innovation. The participants heard talks by entrepreneurs who started out at CERN, sharing their experiences and difficulties.CERN is a hot-bed of high-tech skills and know-how, and the Organization works actively to transfer this expertise to society. Some such innovations can lead to new business start-ups, but it can be extremely difficult to obtain the information and support you need to find your way through the inevitable administrative labyrinth. By opening its doors to the Fondation franco-suisse pour la recherche et la technologie (FFSRT) and hosting this "Entrepreneurial Day" on 7 March, CERN has clearly flagged its desire to assist budding entrepreneurs. The day was jointly kicked off by Olga Hooft, General Manager of the FFSRT, and Maximilian Metzger, CERN’s Se...

  3. Material challenges for the next generation of fission reactor systems

    International Nuclear Information System (INIS)

    Buckthorpe, Derek

    2010-01-01

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO 2 emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  4. Material challenges for the next generation of fission reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Buckthorpe, Derek [AMEC, Knutsford, Cheshire (United Kingdom)

    2010-07-01

    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO{sub 2} emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  5. Ageing significantly at Temelin start-up process

    International Nuclear Information System (INIS)

    Brumovsky, M.; Zdarek, J.; Hahn, J.

    2002-01-01

    Full text: NPP Temelin is a unique case in Plant Life Management Programme application as it is now only in the period of start-up and thus all measures can be easily prepared and implemented in advance based on the experience of similar plants already in operation. Detailed analyses of potential damaging mechanisms as well as of initial conditions and properties of main components have been already done. Regarding this analysis, several measures have been prepared and put into operation: List of important parameters to be measured, saved and periodically analysed have been prepared - some of these parameters are measured in the framework of standard I and C system; new sets of sensors were installed for some additional ones; Special software for on-line evaluation of stress/temperature conditions and subsequently also fatigue damage in some chosen parts (with flow stratification etc.) have been installed and during reactor start-up tests have been checked; Surveillance specimen programme for reactor pressure vessel material monitoring was modified for better and fuller characterisation of material changes during operation; Ex-vessel neutron flux measurement has been established as a part of surveillance specimen programme. The whole Plant Life Management Programme is governed by a special PLIM Group in the plant with necessary links and responsibilities and rules with other necessary plant departments. (author)

  6. Thermoradiation treatment of sewage sludge using reactor waste fission products

    International Nuclear Information System (INIS)

    Reynolds, M.C.; Hagengruber, R.L.; Zuppero, A.C.

    1974-06-01

    The hazards to public health associated with the application of municipal sewage sludge to land usage are reviewed to establish the need for disinfection of sludge prior to its distribution as a fertilizer, especially in the production of food and fodder. The use of ionizing radiation in conjunction with mild heating is shown to be an effective disinfection treatment and an economical one when reactor waste fission products are utilized. A program for researching and experimental demonstration of the process on sludges is also outlined

  7. Gas dynamics models for an oscillating gaseous core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Dam, H. van; Hoogenboom, J.E. (Interuniversitair Reactor Inst., Delft (Netherlands))

    1991-01-01

    Two one-dimensional models are developed for the investigation of the gas dynamical behaviour of the fuel gas in a cylindrical gaseous core fission reactor. By numerical and analytical calculations, it is shown that, for the case where a direct energy extraction mechanism (such as magneto-hydrodynamics (MHD)) is not present, increasing density oscillations occur in the gas. Also an estimate is made of the attainable direct energy conversion efficiency, for the case where a direct energy extraction mechanism is present. (author).

  8. Nuclear data requirements for fission reactor neutronics calculations

    International Nuclear Information System (INIS)

    Finck, P.

    1998-01-01

    The paper discusses current European nuclear data measurement and evaluation requirements for fission reactor technology applications and problems involved in meeting the requirements. Reference is made to the NEA High Priority Nuclear Data Request List and to the production of the new JEFF-3 library of evaluated nuclear data. There are requirements for both differential (or basic) nuclear data measurements and for different types of integral measurement critical facility measurements and isotopic sample irradiation measurements. Cross-section adjustment procedures are being used to take into account the simpler types of integral measurement, and to define accuracy needs for evaluated nuclear data

  9. Review of the neutron capture process in fission reactors

    International Nuclear Information System (INIS)

    Poenitz, W.P.

    1981-07-01

    The importance of the neutron capture process and the status of the more important cross section data are reviewed. The capture in fertile and fissile nuclei is considered. For thermal reactors the thermal to epithermal capture ratio for 238 U and 232 Th remains a problem though some improvements were made with more recent measurements. The capture cross section of 238 U in the fast energy range remains quite uncertain and a long standing discrepancy for the calculated versus experimental central reaction rate ratio C28/F49 persists. Capture in structural materials, fission product nuclei and the higher actinides is also considered

  10. Preparation of mandatory documentation before the start up of the RA-0 `zero power` nuclear reactor at Cordoba National University; Preparacion de la documentacion mandatoria para la puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Martin, H R; Keil, W M; Pezzi, N

    1992-12-31

    Before the start up of the RA-0 `zero power` nuclear reactor installed at Cordoba National University, it was necessary to send to the Regulatory Authority the mandatory documentation which is required in the licensing process. With the previous papers existing for the operation in the first years of the `70, a work program for the future operational training personnel was elaborated. Based on the Authority`s applicable rules and the recommendations and with particular criteria originated in the working university conditions, the SAFETY report of RA-0 nuclear reactor was prepared. This paper describes the principal contents, items and documents involved in the safety report. (Author). [Espanol] Con motivo de la nueva puesta en servicio del REACTOR NUCLEAR RA-0 fue necesario elaborar la documentacion mandatoria requerida por la Autoridad Regulatoria Nacional. Siguiendo los lineamientos de las normas y recomendaciones vigentes e incluyendo criterios propios en lo que debia ser el contenido final de dicha documentacion, fue preparado lo que se ha denominado el INFORME DE SEGURIDAD DEL REACTOR NUCLEAR RA-0. Este documento que se describe en este trabajo, si bien contiene las habituales descripciones de todos los Informes de Seguridad, incluye otros aspectos que no siendo requeridos expresamente en el mismo, han dado una mayor coherencia a la conformacion de todos los aspectos que interrelacionan las areas de seguridad fisica, radiologica, nuclear y de control de materiales nucleares bajo salvaguardias. (Autor).

  11. Hydrogen energy network start-up scenario

    International Nuclear Information System (INIS)

    Weingartner, S.; Ellerbrock, H.

    1994-01-01

    Hydrogen is widely discussed as future fuel and energy storage medium either to replace conventional fuels for automobiles, aircrafts and ships or to avoid the necessity of bulky battery systems for electricity storage, especially in connection with solar power systems. These discussions however started more than 25 years ago and up to now hydrogen has failed to achieve a major break-through towards wider application as energy storage medium in civil markets. The main reason is that other fuels are cheaper and very well implemented in our daily life. A study has been performed at Deutsche Aerospace in order to evaluate the boundary conditions, either political or economical, which would give hydrogen the necessary push, i.e. advantage over conventional fuels. The main goal of this study was to identify critical influence factors and specific start-up scenarios which would allow an economical and practically realistic use of hydrogen as fuel and energy medium in certain niche markets outside the space industry. Method and major results of this study are presented in detail in the paper. Certain niche markets could be identified, where with little initial governmental support, either by funding, tax laws or legislation, hydrogen can compete with conventional fuels. This however requires a scenario where a lot of small actions have to be taken by a high variety of institutions and industries which today are not interconnected with each other, i.e. it requires a new cooperative and proactive network between e.g. energy utilities, car industries, those who have a sound experience with hydrogen (space industry, chemical industry) and last, but certainly not the least, the government. Based on the developed scenario precise recommendations are drawn as conclusions

  12. Starting up microbial enhanced oil recovery.

    Science.gov (United States)

    Siegert, Michael; Sitte, Jana; Galushko, Alexander; Krüger, Martin

    2014-01-01

    This chapter gives the reader a practical introduction into microbial enhanced oil recovery (MEOR) including the microbial production of natural gas from oil. Decision makers who consider the use of one of these technologies are provided with the required scientific background as well as with practical advice for upgrading an existing laboratory in order to conduct microbiological experiments. We believe that the conversion of residual oil into natural gas (methane) and the in situ production of biosurfactants are the most promising approaches for MEOR and therefore focus on these topics. Moreover, we give an introduction to the microbiology of oilfields and demonstrate that in situ microorganisms as well as injected cultures can help displace unrecoverable oil in place (OIP). After an initial research phase, the enhanced oil recovery (EOR) manager must decide whether MEOR would be economical. MEOR generally improves oil production but the increment may not justify the investment. Therefore, we provide a brief economical assessment at the end of this chapter. We describe the necessary state-of-the-art scientific equipment to guide EOR managers towards an appropriate MEOR strategy. Because it is inevitable to characterize the microbial community of an oilfield that should be treated using MEOR techniques, we describe three complementary start-up approaches. These are: (i) culturing methods, (ii) the characterization of microbial communities and possible bio-geochemical pathways by using molecular biology methods, and (iii) interfacial tension measurements. In conclusion, we hope that this chapter will facilitate a decision on whether to launch MEOR activities. We also provide an update on relevant literature for experienced MEOR researchers and oilfield operators. Microbiologists will learn about basic principles of interface physics needed to study the impact of microorganisms living on oil droplets. Last but not least, students and technicians trying to understand

  13. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

    2011-07-01

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

  14. A proposed standard on medical isotope production in fission reactors

    International Nuclear Information System (INIS)

    Schenter, R. E.; Brown, G. J.; Holden, C. S.

    2006-01-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  15. Status report about the works for the start up of the RA-0 `zero power` nuclear reactor at the Cordoba National University; Estado actual de avance de las tareas para la nueva puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Martin, H R; Carballido, C; Oliveras, T

    1992-12-31

    After two years of works at the Cordoba National University for the new start-up of the RA-0 `zero power` nuclear reactor, the results obtained are herein presented. Starting with practically null infrastructure at the beginning, specially in human resources and instrumentation of the reactor, the objectives can be considered satisfactory. The training in work of the future operational staff, the design and the construction of the instrumentation and the fitting of the installations are the principal items described in this paper. An special attention is devoted to the insertion of this type of installation in the university organization, usually not prepared for the quality and control activities, which is necessarily considered in these type of works. (Author). [Espanol] Luego de aproximadamente dos anos de trabajo para la nueva puesta en marcha del REACTOR NUCLEAR RA-0, se han alcanzado los resultados presentados en este trabajo. Partiendo de una infraestructura practicamente inexistente en cuanto a recursos humanos y estado de las instalaciones, los avances logrados son significativos. Comenzando por la capacitacion y el entrenamiento del futuro personal de operacion y pasando por la adecuacion de los equipos y componentes, hasta la confeccion de la documentacion mandatoria, se muestran los aspectos mas destacables de los trabajos realizados. Una atencion especial se dedica a la insercion de una instalacion de este tipo en el ambito universitario, el cual por sus particulares caracteristicas, ha debido ser tenido en cuenta permanentemente para la futura operacion de las instalaciones. (Autor).

  16. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    International Nuclear Information System (INIS)

    Was, Gary S.

    2007-01-01

    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems

  17. Start-up of the ohmic phase in JET

    International Nuclear Information System (INIS)

    Tanga, A.; Christiansen, J.P.; Cordey, J.G.; Ejima, S.; Kellman, A.; Lazzaro, E.; Lomas, P.J.; Thomas, P.R.

    1985-01-01

    JET has been designed to permit the study of plasmas in which alphaparticle heating is a significant part of the power balance. In order to have a sufficient thermonuclear yield and to trap the resulting alphaparticles, JET is similar in its dimensions and plasma current to the next generation of reactor-like devices such as NET, FER and INTOR. For this reason, the authors see the results from the study of the start-up of ohmically heated plasmas in JET as highly relevant. Discussed is the range that has been achieved in all major parameters with ohmic heating. Experiences with the wall conditioning technique and the results of ion cyclotron heating experiments in JET are outlined. This paper also describes the stages of plasma formation, current rise and ohmic flat-top

  18. Start-up of Rapsodie; Le demarrage de Rapsodie

    Energy Technology Data Exchange (ETDEWEB)

    Pontier, R [Commissariat a l' Energie Atomique, 13 - Cadarache (France). Centre d' Etudes Nucleaires

    1967-07-01

    After giving a general description of Rapsodie this report presents the conditions in which the start-up occurred and in which the tests were carried out. A chronological account is given of the operations and of the main events which occurred. The modifications made to the reactor during this period are described and a synthesis of the results obtained is presented. (author) [French] Apres avoir donne de RAPSODIE une description tres generale, ce rapport expose le cadre dans lequel ont ete effectues le demarrage et les essais de l'installation, fait un historique du deroulement des operations et des principaux evenements qui sont intervenus, indique les modifications apportees au reacteur au cours de cette periode et presente une synthese des resultats obtenus. (auteur)

  19. Muon catalyzed fusion - fission reactor driven by a recirculating beam

    International Nuclear Information System (INIS)

    Eliezer, S.; Tajima, T.; Rosenbluth, M.N.

    1986-01-01

    The recent experimentally inferred value of multiplicity of fusion of deuterium and tritium catalyzed by muons has rekindled interest in its application to reactors. Since the main energy expended is in pion (and consequent muon) productions, we try to minimize the pion loss by magnetically confining pions where they are created. Although it appears at this moment not possible to achieve energy gain by pure fusion, it is possible to gain energy by combining catalyzed fusion with fission blankets. We present two new ideas that improve the muon fusion reactor concept. The first idea is to combine the target, the converter of pions into muons, and the synthesizer into one (the synergetic concept). This is accomplished by injecting a tritium or deuterium beam of 1 GeV/nucleon into DT fuel contained in a magnetic mirror. The confined pions slow down and decay into muons, which are confined in the fuel causing little muon loss. The necessary quantity of tritium to keep the reactor viable has been derived. The second idea is that the beam passing through the target is collected for reuse and recirculated, while the strongly interacted portion of the beam is directed to electronuclear blankets. The present concepts are based on known technologies and on known physical processes and data. 29 refs., 6 figs., 4 tabs

  20. Comparison between MBR and SBR on Anammox start-up process from the conventional activated sludge.

    Science.gov (United States)

    Wang, Tao; Zhang, Hanmin; Gao, Dawen; Yang, Fenglin; Zhang, Guangyi

    2012-10-01

    Anammox start-up performances from the conventional activated sludge were compared between a MBR and SBR. Both the reactors successfully started up Anammox process. The start-up period in the MBR (59 days) was notably shorter than that in the SBR (101 days), and the max nitrogen (NH(4)(+)+NO(2)(-)) removal capacity of 345.2 mg N L(-1) d(-1) in the MBR was also higher than that of 292.0 mg N L(-1) d(-1) in the SBR. FISH analysis showed that Anammox bacteria predominated in both reactors. Phylogenetic analysis further disclosed that the MBR had the better biodiversity of Anammox bacteria and gained a higher ecological stability. Generally, the results showed that MBR exhibited a more excellent performance for Anammox start-up. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

  1. Utilization of fast reactor excess neutrons for burning long-lived fission products

    International Nuclear Information System (INIS)

    Kawashima, K.; Kobayashi, K.; Kaneto, K.

    1995-01-01

    An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

  2. Materials compatibility considerations for a fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    DeVan, J.H.; Tortorelli, P.F.

    1983-01-01

    The Tandem Mirror Hybrid Reactor is a fusion reactor concept that incorporates a fission-suppressed breeding blanket for the production of 233 U to be used in conventional fission power reactors. The present paper reports on compatibility considerations related to the blanket design. These considerations include solid-solid interactions and liquid metal corrosion. Potential problems are discussed relative to the reference blanket operating temperature (490 0 C) and the recycling time of breeding materials (<1 year)

  3. An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

  4. Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

    Science.gov (United States)

    Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

    2017-12-01

    A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

  5. Utilisation and performance of sodium instrumentation during start-up and initial operation of Phenix

    International Nuclear Information System (INIS)

    Lions, N.; Buis, H.; Baron, J.; Fournier, C.; Gourdon, J.

    1976-01-01

    The main process-instruments on the Phenix reactor are presented with the exception of the FFDL System and of the hydrogen-detector which are described in other papers. The results obtained during reactor start-up and during initial operation of the nuclear power-station are given [fr

  6. Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

    International Nuclear Information System (INIS)

    Ma Xubo; Chen Yixue; Gao Bin

    2013-01-01

    The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

  7. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  8. Some safety studies for conceptual fusion--fission hybrid reactors. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Okrent, D.

    1978-07-01

    The objective of this study was to make a preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors. The study and subsequent analysis was largely based upon reference to one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The blanket is a fast-spectrum, uranium carbide, helium cooled, subcritical reactor, optimized for the production of fissile fuel. An attempt was made to generalize the results wherever possible

  9. Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

    CERN Document Server

    Kessler, Günter

    2012-01-01

    Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

  10. Environmental mitigation for SCC initiation of BWR core internals by hydrogen injection during start-up

    International Nuclear Information System (INIS)

    Dozaki, K.; Abe, A.; Nagata, N.; Takiguchi, H.

    2004-01-01

    Hydrogen injection into the reactor water has been applied to many BWR power stations. Since hydrogen injected accelerates recombination of oxidant generated by water radiolysis, oxidant concentration, such as dissolved oxygen concentration in reactor water can be reduced. As the result of the reduction of oxidant concentration, Electrochemical Corrosion Potential (ECP) at the surface of structural material can be lowered. Lowered ECP moderates Stress Corrosion Cracking (SCC) sensitivity of structural materials, such as stainless steels. As usual, hydrogen injection system begins to work after the plant start-up is finished, when the condition of normal operation is established. Accordingly, Hydrogen Water Chemistry (HWC) does not cover all the period of plant operation. As far as SCC crack growth is considered, loss of HWC during plant start-up does not result in significant crack growth, because of duration of plant start-up is much shorter than that of plant normal operation, when HWC condition is being satisfied. However, the reactor water environment and load conditions during a plant start-up may contribute to the initiation of SCC. It is estimated that the core internals are subjected to the strain rate that may cause susceptibility to SCC initiation during start-up. Dissolved oxygen (DO) and hydrogen peroxide (H 2 O 2 ) has a peak, and ECP is in high levels during start-up. Therefore it is beneficial to perform hydrogen injection during start-up as well in order to suppress SCC initiation. We call it HWC During Start-up (HDS) here. (orig.)

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

    International Nuclear Information System (INIS)

    Schikorr, W.M.

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

  12. Study of advanced fission power reactor development for the United States. Volume I

    International Nuclear Information System (INIS)

    1976-01-01

    This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

  13. Energy distribution of antineutrinos originating from the decay of fission products in a nuclear reactor

    International Nuclear Information System (INIS)

    Rudstam, G.; Aleklett, K.

    1979-01-01

    The energy spectrum of antineutrinos around a nuclear reactor has been derived by summing contributions from individual fission products. The resulting spectrum is weaker at energies above approx. 8 MeV than earlier published antineutrino spectra. The reason may be connected to the strong feeding of high-lying daughter states in the beta decay of fission products with high disintegration energies

  14. Calculation of burnup and power dependence on fission gas released from PWR type reactor fuel element

    International Nuclear Information System (INIS)

    Edy-Sulistyono

    1996-01-01

    Burn up dependence of fission gas released and variation power analysis have been conducted using FEMXI-IV computer code program for Pressure Water Reactor Fuel During steady-state condition. The analysis result shows that the fission gas release is sensitive to the fuel temperature, the increasing of burn up and power in the fuel element under irradiation experiment

  15. Comparative energetics of three fusion-fission symbiotic nuclear reactor systems

    International Nuclear Information System (INIS)

    Gordon, C.W.; Harms, A.A.

    1975-01-01

    The energetics of three symbiotic fusion-fission nuclear reactor concepts are investigated. The fuel and power balances are considered for various values of systems parameters. The results from this analysis suggest that symbiotic fusion-fission systems are advantageous from the standpoint of economy and resource utilization. (Auth.)

  16. Commissioning and start-up of RA-8 critical assembly

    International Nuclear Information System (INIS)

    Lorenzo, N. de; Diaz, C.; Facchini, G.; Fernandez, C.; Fittipaldi, A.; Juracich, R.; Levanon, I.; Manceda, J.; Martinez, J.; Mogdan, R.; Perez, J.; Scarnichia, E.; Blaumann, H.; Gennuso, G.; Scotti, G.

    1999-01-01

    The RA-8 critical assembly was designed as one of the experimental facilities for the CAREM Reactor Project. This paper describes the activities developed during the cold and hot commissioning, pointing out the difficulties and the solutions applied (some of them original ones). Moreover, this paper will show the main features of the newest nuclear installation of CNEA making a brief description of its characteristics. Among the special circumstances related to the commissioning that are described in the paper we can mention the following: 1. The facility shares the building with the Thermohydraulic Assay Laboratory (L.E.T.), another experimental facility of CAREM, and thus some shared systems have already been working for many years before this start up. Special procedures for these systems were designed to verify the proper functioning under the new requirements. 2. A new driving mechanism, based in hydraulic cylinders, was used to move the control rods. The criteria for acceptance and a validation of the procedure completeness have been carried out. 3. The implementation of a power measurement system based in neutron noise. 4. Measurement of Power Distribution using direct gamma counting from the fuel elements. 5. The commissioning was interrupted for a ten-month period because the personnel involved had to carry out the commissioning of the Egyptian Research Reactor 2. Also, the common activities during a commissioning are described, pointing out the major steps carried out and the results obtained. The following are examples of these activities: 1. Environmental dose survey (before fuel loading and during other stages). 2. Test of equipment and systems isolated from the rest of the plant. 3. Integrated system test (two or more systems working at the same time). 4. Start-up and power operation simulations before fuel loading. 5. Fuel loading strategy during the approximation to criticality by mass. 6. Modification of systems' components to improve the

  17. Conceptual design of a fusion-fission hybrid reactor for transmutation of high level nuclear waste

    International Nuclear Information System (INIS)

    Qiu, L.J.; Wu, Y.C.; Yang, Y.W.; Wu, Y.; Luan, G.S.; Xu, Q.; Guo, Z.J.; Xiao, B.J.

    1994-01-01

    To assess the feasibility of the transmutation of long-lived radioactive waste using fusion-fission hybrid reactors, we are studying all the possible types of blanket, including a comparison of the thermal and fast neutron spectrum blankets. Conceptual designs of a small tokamak hybrid blanket with small inventory of actinides and fission products are presented. The small inventory of wastes makes the system safer. The small hybrid reactor system based on a fusion core with experimental parameters to be realized in the near future can effectively transmute actinides and fission products at a neutron wall loading of 1MWm -2 . An innovative energy system is also presented, including a fusion driver, fuel breeder, high level waste transmuter, fission reactor and so on. An optimal combination of all types of reactor is proposed in the system. ((orig.))

  18. Start-up support for New Brunswick Electric's Point Lepreau nuclear steam generators

    International Nuclear Information System (INIS)

    Schneider, W.; Leroux, A.

    1983-05-01

    The start-up of the 600 MW Point Lepreau reactor provided the opportunity for direct involvement in the important low and medium power start-up phase which was of particular interest because this was a first-of-a-kind reactor type incorporating a new steam generator design. Support included test assistance and test results interpretation for the thermal hydraulic performance of the steam generators and in particular, investigation of water level response to operating pressure, power and feed flow. This work resulted in both a greatly improved understanding of transient characteristics and in a number of beneficial refinements in the control methods

  19. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    1985-01-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  20. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity.

  1. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    Energy Technology Data Exchange (ETDEWEB)

    Asner, D.M.; Burns, K.; Campbell, L.W.; Greenfield, B.; Kos, M.S., E-mail: markskos@gmail.com; Orrell, J.L.; Schram, M.; VanDevender, B.; Wood, L.S.; Wootan, D.W.

    2015-03-11

    The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

  2. Transport and release of fission products during nuclear reactor accident

    International Nuclear Information System (INIS)

    Lee, K.W.; Kuhlman, M.R.; Gieseke, J.A.

    1984-01-01

    This study represents the identification and formulation of a systematic, mechanistic approach to estimating source terms and the implementation of this approach through calculations of fission products release to the environment for a large PWR reactor under a selected set of accident conditions. The development and improvement of calculational procedures is an evolutionary process and in the long term must be verified through experimental studies. It is anticipated that as additional information is obtained the accuracy of predictions can be improved and uncertainties reduced. Transport and deposition of radionuclides were found to be quite dependent on the accident sequences and the corresponding thremal hydraulic conditions. Reduced temperatures led to increased deposition of vapor species, and reduced flow rates to increased aerosol deposition. It is to be recognized that the estimates of release fractions are subject to uncertainties in the data and computer models employed in the calculations and are expected to have been influenced by assumptions regarding plant geometry, thermal hydraulics, deposition mechanisms, and sequence events. The effects of these assumptions will be investigated as this study continues. (Author)

  3. Cerenkov Detectors for Fission Product Monitoring in Reactor Coolant Water

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O

    1967-09-15

    The expected properties of Cerenkov detectors when used for fission product monitoring in water cooled reactors and test loops are discussed from the point of view of the knowledge of the sensitivity of these detectors to some beta emitting isotopes. The basic theory for calculation of the detector response is presented, taking the optical transmission in the sample container and the properties of the photomultiplier tube into account. Special attention is paid to the energy resolution of this type of Cerenkov detector. For the design of practical detectors the results from several investigations of various window and reflector materials are given, and the selection of photomultiplier tubes is briefly discussed. In the case of optical reflectors and photomultiplier tubes reference is made to two previous reports by the author. The influence of the size and geometry of the sample container on the energy resolution follows from a separate investigation, as well as the relative merits of sample containers with transparent inner walls. Provided that the energy resolution of the Cerenkov detector is sufficiently high, there are several reasons for using this detector type for failed-fuel-element detection. It seems possible to attain the desired energy resolution by careful detector design.

  4. Advanced nuclear fuel production by using fission-fusion hybrid reactor

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.; Sahin, S.; Abdulraoof, M.

    1993-01-01

    Efforts are made at the College of Engineering, King Saud University, Riyadh to lay out the main structure of a prototype experimental fusion and fusion-fission (hybrid) reactor blanket in cylindrical geometry. The geometry is consistent with most of the current fusion and hybrid reactor design concepts in respect of the neutronic considerations. Characteristics of the fusion chamber, fusion neutrons and the blanket are provided. The studies have further shown that 1 GWe fission-fusion reactor can produce up to 957 kg/year which is enough to fuel five light water reactors of comparable power. Fuel production can be increased further. 29 refs

  5. Start-up Costs, Taxes and Innovative Entrepreneurship

    NARCIS (Netherlands)

    P. Darnihamedani (Pourya); J.H. Block (Jörn); S.J.A. Hessels (Jolanda); A. Simonyan (Aram)

    2015-01-01

    markdownabstract__Abstract__ Prior research suggests that start-up costs and taxes negatively influence entry into entrepreneurship. Yet, no distinction is made regarding the type of entrepreneurship, particularly innovative versus non-innovative entrepreneurship. Start-up costs, being one-off

  6. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  7. Determination of the fission coefficients in thermal nuclear reactors for antineutrino detection

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Lenilson M. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Cabral, Ronaldo G., E-mail: rgcabral@ime.eb.b [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Anjos, Joao C.C. dos, E-mail: janjos@cbpf.b [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil). Dept. GLN - G

    2011-07-01

    The nuclear reactors in operation periodically need to change their fuel. It is during this process that these reactors are more vulnerable to occurring of several situations of fuel diversion, thus the monitoring of the nuclear installations is indispensable to avoid events of this nature. Considering this fact, the most promissory technique to be used for the nuclear safeguard for the nonproliferation of nuclear weapons, it is based on the detection and spectroscopy of antineutrino from fissions that occur in the nuclear reactors. The detection and spectroscopy of antineutrino, they both depend on the single contribution for the total number of fission of each actinide in the core reactor, these contributions receive the name of fission coefficients. The goal of this research is to show the computational and mathematical modeling used to determinate these coefficients for PWR reactors. (author)

  8. Role of organic matter in the Proterozoic Oklo natural fission reactors, Gabon, Africa

    International Nuclear Information System (INIS)

    Nagy, B.; Rigali, M.J.; Gauthier-Lafaye, F.; Holliger, P.; Mossman, D.J.; Leventhal, J.S.

    1993-01-01

    Of the sixteen known Oklo and the Bangombe natural fission reactors (hydrothermally altered elastic sedimentary rocks that contain abundant uraninite and authigenic clay minerals), reactors 1 to 6 at Oklo contain only traces of organic matter, but the others are rich in organic substances. Reactors 7 to 9 are the subjects of this study. These organic-rich reactors may serve as time-tested analogues for anthropogenic nuclear-waste containment strategies. Organic matter helped to concentrate quantities of uranium sufficient to initiate the nuclear chain reactions. Liquid bitumen was generated from organic matter by hydrothermal reactions during nuclear criticality. The bitumen soon became a solid, consisting of polycyclic aromatic hydrocarbons and an intimate mixture of cryptocrystalline graphite, which enclosed and immobilized uraninite and the fission-generated isotopes entrapped in uraninite. This mechanism prevented major loss of uranium and fission products from the natural nuclear reactors for 1.2 b.y. 24 refs., 4 figs

  9. System model for analysis of the mirror fusion-fission reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Carlson, G.A.

    1977-01-01

    This report describes a system model for the mirror fusion-fission reactor. In this model we include a reactor description as well as analyses of capital cost and blanket fuel management. In addition, we provide an economic analysis evaluating the cost of producing the two hybrid products, fissile fuel and electricity. We also furnish the results of a limited parametric analysis of the modeled reactor, illustrating the technological and economic implications of varying some important reactor design parameters

  10. Preliminary analyses for HTTR`s start-up physics tests by Monte Carlo code MVP

    Energy Technology Data Exchange (ETDEWEB)

    Nojiri, Naoki [Science and Technology Agency, Tokyo (Japan); Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  11. Preliminary analyses for HTTR's start-up physics tests by Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  12. Impurity control and its impact upon start-up and transformer recharging in NET

    International Nuclear Information System (INIS)

    Harrison, M.F.A.

    1985-01-01

    Control of the release of impurities and their subsequent ingress and exhaust from tokamak plasmas has been the subject of intensive studies aimed at both the prediction of reactor burn condition and the interpretation of results from present experiments. Control concepts which are specific to current-initiation, current ramp-up and RF heating during start-up of a reactor such as NET have to date received little attention, according to the author. This paper presents an inductive start-up scenario which is typical of those presently being considered for NET. Shown are plasma configurations during start-up. Particularly significant issues are the advantages of forming the divertor configuration at the earliest practicable stage and the problems expected due to contamination of the limiter with divertor material

  13. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  14. Health effects of low dose exposure to fission products from Chernobyl and the Fermi nuclear reactor in the population of the Detroit metropolitan area

    Energy Technology Data Exchange (ETDEWEB)

    Sternglass, E.J. [Dept. of Radiology, Pittsburgh Univ. School of Medicine, PA (United States); Mangano, J.J.; Gould, J.M. [Radiation and Public Health Project, New York, NY (United States)

    2001-07-01

    The present paper describes the results of the exposure of a very large population in the Detroit, Michigan, area to fallout from Chernobyl measured in 1986, followed by the reported releases from the start-up of the Fermi-II nuclear plant in 1988 located 20 miles from the city that receives its drinking water from Lake St. Clair downwind to the north-east of the plant. Due to the prior existence of a local cancer registry for a total population of about 4 million, and the availability of reliable public-heath statistics by age, race and sex, combined with the absence of an accident known to produce population movement and stress, highly significant rises and declines of the incidence of early childhood leukemia and other cancers could be related both geographically and temporally to the observed rises and declines of fission products in the milk as well as releases from the reactor. Furthermore, surprisingly rapid rises in the incidence of breast cancer also took place in Monroe County where the reactor is located and in Macomb County downwind on Lake St. Clair to the northeast, presumably due to weakening of the immune defenses by the mix of fission products not seen so rapidly after exposure in the case of external X-rays or gamma rays. For Michigan as a whole, for which incidence of thyroid cancer at all ages combined became available after 1985, rapid rises were observed after Chernobyl and the start of the Fermi plant, using as rapidly as in the case of Belarus and Connecticut. Additionally, highly significant synchronous rises in low birth weight, infant mortality, fetal deaths, asthma and infectious disease mortality were also observed consistent with the known action of bone-seeking fission products on the immune system, following reported nuclear tests, nuclear accidents and the start-up of the Fermi plant. (orig.)

  15. Health effects of low dose exposure to fission products from Chernobyl and the Fermi nuclear reactor in the population of the Detroit metropolitan area

    International Nuclear Information System (INIS)

    Sternglass, E.J.; Mangano, J.J.; Gould, J.M.

    2001-01-01

    The present paper describes the results of the exposure of a very large population in the Detroit, Michigan, area to fallout from Chernobyl measured in 1986, followed by the reported releases from the start-up of the Fermi-II nuclear plant in 1988 located 20 miles from the city that receives its drinking water from Lake St. Clair downwind to the north-east of the plant. Due to the prior existence of a local cancer registry for a total population of about 4 million, and the availability of reliable public-heath statistics by age, race and sex, combined with the absence of an accident known to produce population movement and stress, highly significant rises and declines of the incidence of early childhood leukemia and other cancers could be related both geographically and temporally to the observed rises and declines of fission products in the milk as well as releases from the reactor. Furthermore, surprisingly rapid rises in the incidence of breast cancer also took place in Monroe County where the reactor is located and in Macomb County downwind on Lake St. Clair to the northeast, presumably due to weakening of the immune defenses by the mix of fission products not seen so rapidly after exposure in the case of external X-rays or gamma rays. For Michigan as a whole, for which incidence of thyroid cancer at all ages combined became available after 1985, rapid rises were observed after Chernobyl and the start of the Fermi plant, using as rapidly as in the case of Belarus and Connecticut. Additionally, highly significant synchronous rises in low birth weight, infant mortality, fetal deaths, asthma and infectious disease mortality were also observed consistent with the known action of bone-seeking fission products on the immune system, following reported nuclear tests, nuclear accidents and the start-up of the Fermi plant. (orig.)

  16. Tokamak start-up with electron-cyclotron heating

    Energy Technology Data Exchange (ETDEWEB)

    Holly, D J; Prager, S C; Shepard, D A; Sprott, J C [Wisconsin Univ., Madison (USA)

    1981-11-01

    Experiments are described in which the start-up voltage in a tokamak is reduced by about a factor of two by the use of a modest amount of electron cyclotron resonance heating power for pre-ionization. The solution of the zero-dimensional start-up equations indicates that the effect is due to the high initial density which increases the rate at which the conductivity increases in the neutral-dominated initial plasma. The effect extrapolates favourably to larger tokamaks. A 50% reduction in the start-up volt-second requirement and impurity reflux is also observed.

  17. Tokamak start-up with electron-cyclotron heating

    International Nuclear Information System (INIS)

    Holly, D.J.; Prager, S.C.; Shepard, D.A.; Sprott, J.C.

    1981-01-01

    Experiments are described in which the start-up voltage in a tokamak is reduced by about a factor of two by the use of a modest amount of electron cyclotron resonance heating power for pre-ionization. The solution of the zero-dimensional start-up equations indicates that the effect is due to the high initial density which increases the rate at which the conductivity increases in the neutral-dominated initial plasma. The effect extrapolates favourably to larger tokamaks. A 50% reduction in the start-up volt-second requirement and impurity reflux is also observed. (author)

  18. Start-up of NPP Krsko; Pokusno obratovanje NE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Spiler, J; Aralica, J [Nuklearna elektrana Krsko, Krsko (Yugoslavia)

    1984-07-01

    The report describes a review of start-up program and its realisation. There are also described some more significant start-up results with their evaluation. The most significant operation criteria are compared between NPP Krsko and other similar plants in the world. The comparison shows that after the first contractors and operation personnel efforts have been accomplished, our first nuclear power plant is a safe and reliable source of electric power. At the end there are listed NPP Krsko start-up recommendations and experience. (author)

  19. Predicting start-up success with machine learning

    OpenAIRE

    Bento, Francisco Ramadas da Silva Ribeiro

    2018-01-01

    Start-ups are becoming the motor that moves our economy. Google, Apple, or more recently Airbnb and Uber are companies with tremendous impact in worldwide economy, social interactions and government. Over the past decade, both in the US and Europe, there has been an exponential growth in start-up formation. Thus, it seems a relevant challenge understanding what makes this type of high-risk ventures successful and as such, attractive to investors and entrepreneurs. Success for a start-up is de...

  20. Product innovation and commercialization in lean global start-ups

    DEFF Research Database (Denmark)

    Tanev, Stoyan; Rasmussen, Erik Stavnsager; Zijdemans, Erik

    2015-01-01

    The paper examines the distinctive characteristics of product innovation and commercialization in Lean Global Start-up (LGS) – new technology firms which have adopted a lean and global path from or near to their inception. It suggests an uncertainty vs risk framework which allows integrating two...... research streams – Born Global (BG) firms and lean start-ups. In addition to its integrative theoretical value, the paper offers insights for lean start-up managers dealing with the challenges of a global start....

  1. Chooz B1 start-up marked by total computer control

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The start-up of Chooz B1, the first of EdF's latest generation 1450 MWe N4 reactors, marks the first use in a nuclear power plant of a fully computerised control room. Working in partnership with EdF, Sema Group designed and supplied the advanced command and control system for this plant. (Author)

  2. Effect of zero-valent iron and trivalent iron on UASB rapid start-up.

    Science.gov (United States)

    Wang, Jie; Fang, Hongyan; Jia, Hui; Yang, Guang; Gao, Fei; Liu, Wenbin

    2018-01-01

    In order to realize the rapid start-up of upflow anaerobic sludge blanket (UASB) reactor, the iron ion in different valence state was added to UASB. The results indicated that the start-up time of R3 (FeCl 3 ) was 48 h faster than that of R2 (zero-valent iron (ZVI)). It was because the FeCl 3 could rapidly promote granulation of sludge as a flocculant. However, ZVI released Fe 2+ through corrosion slowly, and then the Fe 2+ increased start-up speed by enhancing enzyme activity and enriching methanogens. In addition, the ZVI and FeCl 3 could promote hydrolysis acidification and strengthen the decomposition of long-chain fatty acids. The detection of iron ions showed that iron ions mainly existed in the sludge. Because the high concentration of Fe 2+ could inhibit anaerobic bacteria activity, excess Fe 3+ could be changed into iron hydroxide precipitation to hinder the mass transfer process of anaerobic bacteria under the alkaline condition. The FeCl 3 was suitable to be added at the initial stage of UASB start-up, and the ZVI was more fitted to be used in the middle stage of reactor start-up to improve the redox ability.

  3. Fission Fragment Yield Data in Support of Advanced Reactor Technology

    Energy Technology Data Exchange (ETDEWEB)

    Hecht, Adam [Univ. of New Mexico, Albuquerque, NM (United States)

    2017-11-21

    Within the 3 year POP we propose to continue to test and further develop the fission spectrometers, to do development tests and full data acquisition run at the national laboratory neutron beam facilities, to measure correlated fission fragment yields at low neutron energies with 235 U fission targets, and make these data available to the nuclear community. The spectrometer development will be both on the university based r\\prototype and on the National Laboratory Spectrometer, and measurements will be performed with both. Over the longer time frame of the collaboration, we will take data over a range of low energies, and use other fission targets available to the laboratory. We will gather energy specific fragment distributions and reaction cross sections. We will further develop the data acquisition capabilities to take correlated fission fragment'gamma ray/neurton data, all on an event-by-event basis. This really is an enabling technology.

  4. An evaluation of designed Start-up System (SUS) for once-through steam generators

    International Nuclear Information System (INIS)

    Yu, Dali; Peng, Minjun; Xia, Genglei; Mao, Wanchao; Yang, Yang

    2014-01-01

    Highlights: • Four SUS solutions are established to help reactors complete the start-up process. • Solutions C and D lead to superior behaved than Solutions A and B. • Solution C is fit for propulsion reactors and Solution D for nuclear power reactors. - Abstract: In this paper, we report on research that has been carried out both on the comparison and evaluation of four designed Start-up Systems (SUSs), which are proposed to improve reliability and economy during the reactor start-up process. Firstly, four SUS solutions are established to satisfy demands and detailed descriptions are introduced. Then, four reactor systems, which include SUS, are modeled by the estimate code JTopmeret. Furthermore besides, the JTopmeret code is validated for the once-through steam generators (OTSG). Finally, the behaviors of SUS and the steady and transient performance of OTSG are investigated. The results show that the designed SUS can successfully improve the economy and the OTSG operational safety. Suggestion is made in this paper to apply Solution C in propulsion-purposed reactors, and Solution D in nuclear power reactors

  5. Workshop summaries for the third US/USSR symposium on fusion-fission reactors

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1979-07-01

    Workshop summaries on topics related to the near-term development requirements for fusion-fission (hybrid) reactors are presented. The summary topics are as follows: (1) external factors, (2) plasma engineering, (3) ICF hybrid reactors, (4) blanket design, (5) materials and tritium, and (6) blanket engineering development requirements

  6. CERN confident of LHC start-up in 2007

    CERN Document Server

    2007-01-01

    "Delegates attending the 140th meeting of CERN Council heard a confident report from the Laboratory about the scheduled start-up of the world's highest energy particle accelerator, the Large Hadron Collier (LHC), in 2007." (1 page)

  7. 1-D hybrid code for FRM start-up

    International Nuclear Information System (INIS)

    Stark, R.A.; Miley, G.H.

    1982-01-01

    A one-D hybrid has been developed to study the start-up of the FRM via neutral-beam injection. The code uses a multi-group numerical model originally developed by J. Willenberg to describe fusion product dynamics in a solenoidal plasma. Earlier we described such a model for use in determining self-consistent ion currents and magnetic fields in FRM start-up. However, consideration of electron dynamics during start-up indicate that the electron current will oppose the injected ion current and may even foil the attempt to achieve reversal. For this reason, we have combined the multi-group ion (model) with a fluid treatment for electron dynamics to form the hybrid code FROST (Field Reversed One-dimensional STart-up). The details of this merger, along with sample results of operation of FROST, are given

  8. CERN confirms goal of 2007 start-up for LHC

    CERN Document Server

    2005-01-01

    Speaking at the 131st session of CERN Council on 17 December 2004, the Director-General, Robert Aymar, confirmed that the top priority is to maintain the goal of starting up the Large Hadron Collider (LHC) in 2007.

  9. University Knowledge Spillovers & Regional Start-up Rates

    DEFF Research Database (Denmark)

    Hellerstedt, Karin; Wennberg, Karl; Frederiksen, Lars

    2014-01-01

    how characteristics of the economic and political milieu within each region influence the ratio of firm births. We find that knowledge spillovers from universities and firm-based R&D strongly affect the start-up rates for both high-tech firms and knowledge-intensive services firms. Further, the start......This chapter investigates how regional start-up rates in the knowledge-intensive services and high-tech industries are influenced by knowledge spillovers from both universities and firm-based R&D activities. Integrating insights from economic geography and organizational ecology into the literature......-up rate of knowledge-intensive service firms is tied more strongly to the supply of university educated individuals and the political regulatory regime within the municipality than start-ups in high-tech industries. This suggests that knowledge-intensive service-start-ups are more susceptible to both...

  10. Spring 
start-up for the LHC

    CERN Multimedia

    2007-01-01

    A new schedule for the commissioning of the LHC was presented to the Council at its Session in June. The start-up of the accelerator is scheduled for May 2008, to allow all technical problems to be resolved.

  11. An eigenexpansion technique for modelling plasma start-up

    International Nuclear Information System (INIS)

    Pillsbury, R.D.

    1989-01-01

    An algorithm has been developed and implemented in a computer program that allows the estimation of PF coil voltages required to start-up an axisymmetric plasma in a tokamak in the presence of eddy currents in toroidally continuous conducting structures. The algorithm makes use of an eigen-expansion technique to solve the lumped parameter circuit loop voltage equations associated with the PF coils and passive (conducting) structures. An example of start-up for CIT (Compact Ignition Tokamak) is included

  12. University Start-ups: A Better Business Model

    Science.gov (United States)

    Dehn, J.; Webley, P. W.

    2015-12-01

    Many universities look to start-up companies as a way to attract faculty, supporting research and students as traditional federal sources become harder to come by. University affiliated start-up companies can apply for a broader suite of grants, as well as market their services to a broad customer base. Often university administrators see this as a potential panacea, but national statistics show this is not the case. Rarely do universities profit significantly from their start-ups. With a success rates of around 20%, most start-ups end up costing the university money as well as faculty-time. For the faculty, assuming they want to continue in academia, a start-up is often unattractive because it commonly leads out of academia. Running a successful business as well as maintaining a strong teaching and research load is almost impossible to do at the same time. Most business models and business professionals work outside of academia, and the models taught in business schools do not merge well in a university environment. To mitigate this a new business model is proposed where university start-ups are aligned with the academic and research missions of the university. A university start-up must work within the university, directly support research and students, and the work done maintaining the business be recognized as part of the faculty member's university obligations. This requires a complex conflict of interest management plan and for the companies to be non-profit in order to not jeopardize the university's status. This approach may not work well for all universities, but would be ideal for many to conserve resources and ensure a harmonious relationship with their start-ups and faculty.

  13. Strategic Management: Business Model Canvas for Start-Up Company

    OpenAIRE

    Sonninen, Anna

    2016-01-01

    Thesis is written about development of Kakunpala as a Start-Up Company, because development processes and ideas are the key factors for future success of the company. The aim of this thesis is to understand Strategic Management of the Company and Development processes,by using the SWOT analysis before and after development and by applying Business Model Canvas template to the Start-Up Company, and by describing Customer Journey: what relations Kakunpala has with clients and its processes...

  14. Dampak Mentoring Pada Keberhasilan Start-Up Business: Studi Kasus Pada Start-Up Business di Indonesia [Mentoring the Impact of Success of a Start-Up Business: A Case Study of a Start-Up Business in Indonesia

    OpenAIRE

    Christina Yanita Setyawati

    2016-01-01

    Entrepreneurship is a highly developed science today, as well as in the world of education in Indonesia. Teaching entrepreneurship at the university level, especially at some universities in Surabaya, the second largest city in Indonesia, shows the positive impact that there were 74.03% of start-up businesses that survived and thrived in 2014. Based on the previous observation, some business start-ups are accompanied by a mentor. Mentoring is a process of forming and maintaining lasting relat...

  15. A Review of Previous Research in Direct Energy Conversion Fission Reactors

    International Nuclear Information System (INIS)

    DUONG, HENRY; POLANSKY, GARY F.; SANDERS, THOMAS L.; SIEGEL, MALCOLM D.

    1999-01-01

    From the earliest days of power reactor development, direct energy conversion was an obvious choice to produce high efficiency electric power generation. Directly capturing the energy of the fission fragments produced during nuclear fission avoids the intermediate conversion to thermal energy and the efficiency limitations of classical thermodynamics. Efficiencies of more than 80% are possible, independent of operational temperature. Direct energy conversion fission reactors would possess a number of unique characteristics that would make them very attractive for commercial power generation. These reactors would be modular in design with integral power conversion and operate at low pressures and temperatures. They would operate at high efficiency and produce power well suited for long distance transmission. They would feature large safety margins and passively safe design. Ideally suited to production by advanced manufacturing techniques, direct energy conversion fission reactors could be produced more economically than conventional reactor designs. The history of direct energy conversion can be considered as dating back to 1913 when Moseleyl demonstrated that charged particle emission could be used to buildup a voltage. Soon after the successful operation of a nuclear reactor, E.P. Wigner suggested the use of fission fragments for direct energy conversion. Over a decade after Wigner's suggestion, the first theoretical treatment of the conversion of fission fragment kinetic energy into electrical potential appeared in the literature. Over the ten years that followed, a number of researchers investigated various aspects of fission fragment direct energy conversion. Experiments were performed that validated the basic physics of the concept, but a variety of technical challenges limited the efficiencies that were achieved. Most research in direct energy conversion ceased in the US by the late 1960s. Sporadic interest in the concept appears in the literature until this

  16. AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

  17. Fusion--fission hybrid reactors: a capsule introduction

    International Nuclear Information System (INIS)

    Holdren, J.P.

    1977-01-01

    A short introduction to fusion-fission hybrid systems is provided touching on (a) basic technological characteristics; (b) potential applications; (c) relevance of environmental considerations in the development rationale for hybrids. References to the more technical literature are supplied

  18. Fission product release from fuel of water-cooled reactors

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.; Klisinska, M.

    1997-01-01

    The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

  19. Reference reactor module for NASA's lunar surface fission power system

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  20. Role of fission-reactor-testing capabilities in the development of fusion technology

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

    1981-01-01

    Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

  1. A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

    International Nuclear Information System (INIS)

    Bartram, B.W.; Dougherty, D.K.

    1987-01-01

    This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs

  2. Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

    International Nuclear Information System (INIS)

    Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F.

    2009-01-01

    The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U 235 (typically Pu 242 , Np 237 , U 238 , Th 232 ). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

  3. Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F. [CEA, DEN, Dosimetry Command Control and Instrumentation Laboratory, F-13109 Saint-Paul-lez-Durance (France)

    2009-07-01

    The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U{sup 235} (typically Pu{sup 242}, Np{sup 237}, U{sup 238}, Th{sup 232}). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

  4. The process of the start-up of a PWR nuclear power plant in the USA

    International Nuclear Information System (INIS)

    Rana, B.S.

    1977-01-01

    The procedure is described of putting into full operation the William B. Mc Guire nuclear power plant with two PWR reactors with an output of 2x3,411 MWt (2x1,220 MWe) supplied to the Duke Power Co. lock, stock and barrel. The basic specifications are shown of units I and II and the organization of start-up activities is described. The time schedule of preoperational and start-up tests is shown and testing is reviewed preceding the first fuel charge. Also described are tests related to the first start-up of a unit comprising the period of the first fuel charge, the initial critical state, low-power tests and tests with power gradually increased. In tests of the individual systems and components of the unit, operating regulations are verified and, if required, renewed, or new regulations are introduced. (B.S.)

  5. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Lewis, B.J.

    1983-01-01

    The release of gaseous (krypton and xenon) and iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assumes precursor-corrected 'Booth diffusional release' in the UO 2 and subsequent holdup in the fuel-to-sheath gap. Transport in the gap is separately modelled with a phenomenological rate constant (assuming release from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from possessing various states of defection are use in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small drilled hole. A second element was machined with 23 slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Comparison of measured release data with calculated values from the model yields estimates of empirical diffusion coefficients for the radioactive species in the UO 2 (1.56 x 10 -10 to 7.30 x 10 -9 s -1 ), as well as escape rate constants (7.85 x 10 -7 to 3.44 x 10 -5 s -1 ) and diffusion coefficients (3.39 x 10 -5 to 4.88 x 10 -2 cm 2 /s) for these in the fuel-to-sheath gap. Analyses also enable identification of the various rate-controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is 'Booth diffusion'; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-heath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this work, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size)

  6. Reference and standard benchmark field consensus fission yields for U.S. reactor dosimetry programs

    International Nuclear Information System (INIS)

    Gilliam, D.M.; Helmer, R.G.; Greenwood, R.C.; Rogers, J.W.; Heinrich, R.R.; Popek, R.J.; Kellogg, L.S.; Lippincott, E.P.; Hansen, G.E.; Zimmer, W.H.

    1977-01-01

    Measured fission product yields are reported for three benchmark neutron fields--the BIG-10 fast critical assembly at Los Alamos, the CFRMF fast neutron cavity at INEL, and the thermal column of the NBS Research Reactor. These measurements were carried out by participants in the Interlaboratory LMFBR Reaction Rates (ILRR) program. Fission product generation rates were determined by post-irradiation analysis of gamma-ray emission from fission activation foils. The gamma counting was performed by Ge(Li) spectrometry at INEL, ANL, and HEDL; the sample sent to INEL was also analyzed by NaI(Tl) spectrometry for Ba-140 content. The fission rates were determined by means of the NBS Double Fission Ionization Chamber using thin deposits of each of the fissionable isotopes. Four fissionable isotopes were included in the fast neutron field measurements; these were U-235, U-238, Pu-239, and Np-237. Only U-235 was included in the thermal neutron yield measurements. For the fast neutron fields, consensus yields were determined for three fission product isotopes--Zr-95, Ru-103, and Ba-140. For these fission product isotopes, a separately activated foil was analyzed by each of the three gamma counting laboratories. The experimental standard deviation of the three independent results was typically +- 1.5%. For the thermal neutron field, a consensus value for the Cs-137 yield was also obtained. Subsidiary fission yields are also reported for other isotopes which were studied less intensively (usually by only one of the participating laboratories). Comparisons with EBR-II fast reactor yields from destructive analysis and with ENDF/B recommended values are given

  7. Reaction Rate Benchmark Experiments with Miniature Fission Chambers at the Slovenian TRIGA Mark II Reactor

    Science.gov (United States)

    Štancar, Žiga; Kaiba, Tanja; Snoj, Luka; Barbot, Loïc; Destouches, Christophe; Fourmentel, Damien; Villard, Jean-François AD(; )

    2018-01-01

    A series of fission rate profile measurements with miniature fission chambers, developed by the Commisariat á l'énergie atomique et auxénergies alternatives, were performed at the Jožef Stefan Institute's TRIGA research reactor. Two types of fission chambers with different fissionable coating (235U and 238U) were used to perform axial fission rate profile measurements at various radial positions and several control rod configurations. The experimental campaign was supported by an extensive set of computations, based on a validated Monte Carlo computational model of the TRIGA reactor. The computing effort included neutron transport calculations to support the planning and design of the experiments as well as calculations to aid the evaluation of experimental and computational uncertainties and major biases. The evaluation of uncertainties was performed by employing various types of sensitivity analyses such as experimental parameter perturbation and core reaction rate gradient calculations. It has been found that the experimental uncertainty of the measurements is sufficiently low, i.e. the total relative fission rate uncertainty being approximately 5 %, in order for the experiments to serve as benchmark experiments for validation of fission rate profiles. The effect of the neutron flux redistribution due to the control rod movement was studied by performing measurements and calculations of fission rates and fission chamber responses in different axial and radial positions at different control rod configurations. It was confirmed that the control rod movement affects the position of the maximum in the axial fission rate distribution, as well as the height of the local maxima. The optimal detector position, in which the redistributions would have minimum effect on its signal, was determined.

  8. Solid State Track Recorder fission rate measurements in low power light water reactor pressure vessel mockups

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Kellogg, L.S.

    1985-01-01

    The results of extensive SSTR measurements made at the Pool Critical Assembly (PCA) facility at Oak Ridge National Laboratory have been reported previously. Measurements were made at key locations in PCA which is an idealized mockup of the water gap, thermal shield, pressure vessel geometry of a light water reactor. Recently, additional SSTR fission rate measurements have been carried out for 237-Np, 238-U, and 235-U in key locations in the NESTOR Shielding and Dosimetry Improvement Program (NESDIP) mockup facility located at Winfrith, England. NESDIP is a replica of the PCA facility, and comparisons will be made between PCA and NESDIP measurements. The results of measurements made at the engineering mockup at the VENUS critical assembly at CEN/SCK, Mol, Belgium will also be reported. Measurements were made at selected radial and azimuthal locations in VENUS, which models the in-core and near-core regions of a pressurized water reactor. Comparisons of absolute SSTR fission rates with absolute fission rates made with the Mol miniature fission chamber will be reported. Absolute fission rate comparisons have also been made between the NBS fission chamber, radiometric fission foils, and SSTRs, and these results will be summarized

  9. Preparation of a primary target for the production of fission products in a nuclear reactor

    International Nuclear Information System (INIS)

    Arino, H.; Cosolito, F.J.; George, K.D.; Thornton, A.K.

    1976-01-01

    A primary target for the production of fission products in a nuclear reactor, such as uranium or plutonium fission products, is comprised of an enclosed, cylindrical vessel, preferably comprised of stainless steel, having a thin, continuous, uniform layer of fissionable material, integrally bonded to its inner walls and a port permitting access to the interior of the vessel. A process is also provided for depositing uranium material on to the inner walls of the vessel. Upon irradiation of the target with neutrons from a nuclear reactor, radioactive fission products, such as molybdenum-99, are formed, and thereafter separated from the target by the introduction of an acidic solution through the port to dissolve the irradiated inner layer. The irradiation and dissolution are thus effected in the same vessel without the necessity of transferring the fissionable material and fission products to a separate chemical reactor. Subsequently, the desired isotopes are extracted and purified. Molybdenum-99 decays to technetium-99m which is a valuable medical diagnostic radioisotope. 3 claims, 1 drawing figure

  10. LOFC fission product release and circulating activity calculations for gas-cooled reactors

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

    1977-01-01

    The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

  11. Experience from start-ups of the first ANITA Mox plants.

    Science.gov (United States)

    Christensson, M; Ekström, S; Andersson Chan, A; Le Vaillant, E; Lemaire, R

    2013-01-01

    ANITA™ Mox is a new one-stage deammonification Moving-Bed Biofilm Reactor (MBBR) developed for partial nitrification to nitrite and autotrophic N-removal from N-rich effluents. This deammonification process offers many advantages such as dramatically reduced oxygen requirements, no chemical oxygen demand requirement, lower sludge production, no pre-treatment or requirement of chemicals and thereby being an energy and cost efficient nitrogen removal process. An innovative seeding strategy, the 'BioFarm concept', has been developed in order to decrease the start-up time of new ANITA Mox installations. New ANITA Mox installations are started with typically 3-15% of the added carriers being from the 'BioFarm', with already established anammox biofilm, the rest being new carriers. The first ANITA Mox plant, started up in 2010 at Sjölunda wastewater treatment plant (WWTP) in Malmö, Sweden, proved this seeding concept, reaching an ammonium removal rate of 1.2 kgN/m³ d and approximately 90% ammonia removal within 4 months from start-up. This first ANITA Mox plant is also the BioFarm used for forthcoming installations. Typical features of this first installation were low energy consumption, 1.5 kW/NH4-N-removed, low N₂O emissions, started up at Sundets WWTP in Växjö, Sweden, reached full capacity with more than 90% ammonia removal within 2 months from start-up. By applying a nitrogen loading strategy to the reactor that matches the capacity of the seeding carriers, more than 80% nitrogen removal could be obtained throughout the start-up period.

  12. Start-up Funding via Equity Crowdfunding in Germany

    DEFF Research Database (Denmark)

    Angerer, Martin; Brem, Alexander; Kraus, Sascha

    2017-01-01

    Entrepreneurs often struggle to find sufficient funding for their start-ups. A relatively new way for companies to attract capital is via an internet platform, locating investors who in return receive something in return for their ventures. Equity crowdfunding is one of several types of crowdfund......Entrepreneurs often struggle to find sufficient funding for their start-ups. A relatively new way for companies to attract capital is via an internet platform, locating investors who in return receive something in return for their ventures. Equity crowdfunding is one of several types...... of crowdfunding, and is also known as crowdinvesting in the German-speaking realm. This article predominantly advances the scientific knowledge regarding the success factors of equity crowdfunding for German start-ups. The study conducted nine qualitative interviews with start-ups and crowdinvesting platforms....... Its first result is that German start-ups select crowdinvesting because (1) it is a funding opportunity and (2) it has an expected marketing effect. To organize the results of relevant success factors, the Crowdinvesting Success Model was designed by the researchers. This supports German entrepreneurs...

  13. The Rationality and Irrationality of Financing Green Start-Ups

    Directory of Open Access Journals (Sweden)

    Linda Bergset

    2015-11-01

    Full Text Available Green start-ups contribute towards a transition to a more sustainable economy by developing sustainable and environmentally friendly innovation and bringing it to the market. Due to specific product/service characteristics, entrepreneurial motivation and company strategies that might differ from that of other start-ups, these companies might struggle even more than usual with access to finance in the early stages. This conceptual paper seeks to explain these challenges through the theoretical lenses of entrepreneurial finance and behavioural finance. While entrepreneurial finance theory contributes to a partial understanding of green start-up finance, behavioural finance is able to solve a remaining explanatory deficit produced by entrepreneurial finance theory. Although some behavioural finance theorists are suggesting that the current understanding of economic rationality underlying behavioural finance research is inadequate, most scholars have not yet challenged these assumptions, which constrict a comprehensive and realistic description of the reality of entrepreneurial finance in green start-ups. The aim of the paper is thus, first, to explore the specifics of entrepreneurial finance in green start-ups and, second, to demonstrate the need for a more up-to-date conception of rationality in behavioural finance theory in order to enable realistic empirical research in this field.

  14. Preconceptual design and analysis of a solid-breeder blanket test in an existing fission reactor

    International Nuclear Information System (INIS)

    Deis, G.A.; Hsu, P.Y.; Watts, K.D.

    1983-01-01

    Preconceptual design and analysis have been performed to examine the capabilities of a proposed fission-based test of a water-cooled Li 2 O blanket concept. The mechanical configuration of the test piece is designed to simulate a unit cell of a breeder-outside-tube concept. This test piece will be placed in a fission test reactor, which provides an environment similar to that in a fusion reactor. The neutron/gamma flux from the reactor produces prototypical power density, tritium production rates, and operating temperatures and stresses. Steady-state tritium recovery from the test piece can be attained in short-duration (5-to-6-day) tests. The capabilities of this test indicate that fission-based testing can provide important near-term engineering information to support the development of fusion technology

  15. New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ma, X.B., E-mail: maxb@ncepu.edu.cn; Qiu, R.M.; Chen, Y.X.

    2017-02-15

    Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between {sup 235}U and {sup 239}Pu, the covariance coefficient changes from 0.15 to −0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller. - Highlights: • The covariance coefficients between isotopes vs reactor burnup may change its sign because of two opposite effects. • The relation between fission fraction uncertainty and atomic density are first studied. • A new MC-based method of evaluating the covariance coefficients between isotopes was proposed.

  16. Studies on the properties of hard-spectrum, actinide fissioning reactors. Final report

    International Nuclear Information System (INIS)

    Nelson, J.B.; Prichard, A.W.; Schofield, P.E.; Robinson, A.H.; Spinrad, B.I.

    1980-01-01

    It is technically feasible to construct an operable (e.g., safe and stable) reactor to burn waste actinides rapidly. The heart of the concept is a driver core of EBR-II type, with a central radial target zone in which fuel elements, made entirely of waste actinides are exposed. This target fuel undergoes fission, as a result of which actinides are rapidly destroyed. Although the same result could be achieved in more conventionally designed LWR or LMFBR systems, the fast spectrum reactor does a much more efficient job, by virtue of the fact that in both LWR and LMFBR reactors, actinide fission is preceded by several captures before a fissile nuclide is formed. In the fast spectrum reactor that is called ABR (actinide burning reactor), these neutron captures are short-circuited

  17. Starting-up sequence of the AWEC-60 wind turbine

    International Nuclear Information System (INIS)

    Avia, F.; Cruz, M. de la.

    1991-01-01

    One of the most critical status of the wind turbines operation is the starting-up sequence and the connection to the grid, due to the actuating loads that could be several times the loads during operation at rated conditions. Due to this fact, the control strategy is very important during the starting-up sequence in order to minimize the loads on the machine. For this purpose it is necessary to analyze the behaviour of the wind turbine during that sequence in different wind conditions and machine conditions. This report shows the graphic information about fifty starting-up sequences of the AWEC-60 wind turbine of 60 m. diameter and 1200 kW of rated power, recorded in April 1991 and cut-out wind speed. (author)

  18. Start-up processes’ efficiency of turbine jet engines

    Directory of Open Access Journals (Sweden)

    Jankowski Antoni

    2016-12-01

    Full Text Available The paper presents diagnoses of example start-up processes of drive units built in military aircraft operated in the Dęblin's “School of Eaglets” during the flight. The significance of this process is very important, especially from the point of view of flight safety, especially as it concerns the training aircraft, on which officer cadets-candidates for pilots are trained. The diagnosis of the start-up process was conducted using data from on-board flight recorders recorded during a flight training, whose aim was an emergency launch of particular drive units, and with the help of the so-called phase mapping of the selected operating parameters of the particular flight drive systems. The standard diagram of typical start-up systems was shown, while presenting their base subassemblies. At the end, the obtained results and the main problems that need intervention were commented, as well as further preventive activities were proposed.

  19. ECRH-assisted start-up in ITER

    International Nuclear Information System (INIS)

    Lloyd, B.; Carolan, P.G.; Warrick, C.D.

    1996-07-01

    In ITER, the electric field applied for ionisation and to ramp up the plasma current may be limited to ∼ 0.3 V/m. In this case, based on established theories of the avalanche process, it is shown that ohmic breakdown in ITER is only possible over a narrow range of pressure and magnetic error field. Therefore, ECRH may be necessary to provide robust and reliable start-up. ECRH can ensure prompt breakdown over a wide range of prefill pressure and error field and can also give control over the initial time and location of breakdown. For ECRH-assisted start-up in ITER, the power and pulse length requirements are essentially determined by the need to ensure burnthrough, i.e. complete ionisation of hydrogen and the transition to high ionisation states of impurities. A 0-D code (with inclusion of some 1-D effects) has been developed to analyse burnthrough in ITER. The 0-D simulations indicate that control of the deuterium density is the key factor for ensuring successful start-up in ITER, where the effects of neutral screening and dynamic fuelling by the ex-plasma volume are also crucial. It is concluded that without ECRH, successful start-up will only be possible over a very restricted range of parameters but 3MW of absorbed ECRH power will ensure reasonably robust start-up for a broad range of conditions with beryllium impurity. In the case of carbon impurity, even with an absorbed ECRH power of 5MW one may be restricted to low prefill pressure and/or low carbon concentration for successful start-up. (Author)

  20. Progress on the conceptual design of a mirror hybrid fusion--fission reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Burleigh, R.J.

    1975-01-01

    A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel

  1. The start up as a phase of architectural design process.

    Science.gov (United States)

    Castro, Iara Sousa; Lima, Francisco de Paula Antunes; Duarte, Francisco José de Castro Moura

    2012-01-01

    Alterations made in the architectural design can be considered as a continuous process, from its conception to the moment a built environment is already in use. This article focuses on the "moving phase", which is the initial moment of the environment occupation and the start-up of services. It aims to show that the continuity of ergonomics interventions during the "moving phase" or start up may reveal the built environment inadequacies; clearly showing needs not met by the design and allowing making instant decisions to solve non-foreseen problems. The results have revealed some lessons experienced by users during a critical stage not usually included in the design process.

  2. Female Hires and the Success of Start-up Firms

    OpenAIRE

    Weber, Andrea; Zulehner, Christine

    2009-01-01

    In this paper we investigate the relationship between females among the first hires of start-up companies and business success. Our results show that firms with female first hires have a higher share of female workers at the end of the first year after entry. Further, we find that firms with female first hires are more successful and stay longer in the market. We conclude that our results support the hypothesis that gender-diversity in leading positions is an advantage for start-up firms.

  3. Start-up of plasma current by electron Bernstein wave

    International Nuclear Information System (INIS)

    Maekawa, Takashi; Tanaka, Hitoshi; Uehide, Masaki

    2009-01-01

    Electron cyclotron current drive by electron Bernstein (EB) waves for the start-up and ramp-up of toroidal plasma current with no central solenoid in tokamaks is discussed. It is shown that high N// EB waves have ability to ramp-up the current against the counter voltage from self-induction, where N// is the parallel refractive index to the magnetic field, and they are especially suitable for initial current start-up phase where the bulk electron temperature is low enough to ensure high N// EB waves. (author)

  4. Radiation effects in fuel materials for fission reactors

    International Nuclear Information System (INIS)

    Matzke, H.

    1983-01-01

    Physical and chemical changes that occur in fuel materials during fission are described. Emphasis is placed on the fuels used today, or those foreseen for the future, hence oxides and carbides of uranium and plutonium. Examples are given to illustrate the most interesting neutron effects. (author)

  5. Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

    Directory of Open Access Journals (Sweden)

    Salahuddin Asif

    2013-01-01

    Full Text Available Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

  6. Use of dwell time concept in fission product inventory assessment for CANDU reactors

    International Nuclear Information System (INIS)

    Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

    2003-01-01

    A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

  7. Evaluating effectiveness of project start-ups: an exploratory study

    NARCIS (Netherlands)

    Halman, Johannes I.M.; Burger, G.T.N.

    In this paper an exploratory study is reported about the effectiveness of project start-up (PSU) practices within a world-scale operating, high technology innovating and manufacturing company. The emphasis is on the focal position of both project owner and project manager. To uncover potential

  8. CERN confident of LHC start-up in 2007

    CERN Document Server

    Vanden Broeck, Renilde

    2007-01-01

    "Delegates attending the 140th meeting of CERN Council today heard a confident report from the Laboratory about the scheduled start-up of the world's highest energy particle accelerator, the Large Hadron Collier (LHC) in 2007. (1/2 page)

  9. CERN confident of LHC start-up in 2007

    CERN Multimedia

    2006-01-01

    "Delegates attending the 140th meeting of CERN Council today heard a confident report from the Laboratory about the scheduled start-up of the world's highest energy particle accelerator, the Large Hadron Collider (LHC) in 2007." (1/2 page)

  10. Our start-up could never have been successful

    NARCIS (Netherlands)

    Van Kranendonk, J.; De Leeuw, T.; Bonger, S.

    2014-01-01

    In July 2013, TU Delft alumnus Jan van Kranendonk and his business partner Thomas de Leeuw decided to discontinue their start-up company Sunuru. The reason was that customers were not interested in their ingenious solar project. Van Kranendonk spoke about his failure during the Failcon event in

  11. Social Capital in Knowledge Intensive Start-Ups

    DEFF Research Database (Denmark)

    Neergaard, Helle; Madsen, Henning

    This paper addresses social capital in biotechnology, medico and information communication start-ups using both quantitative and qualitative data. It shows that founding teams are primarily composed of 'trusted alters' and that networking patterns are highly influenced by the entrepreneur...

  12. NEW VENTURE CREATION: HOW START-UPS GROW?

    Directory of Open Access Journals (Sweden)

    AIDIN SALAMZADEH

    Full Text Available ABSTRACT Start-ups, often seen as sources of innovation and change, are prone to failure and accordingly they are attracting considerable attention not least from policy makers and Government officials. However, the various new venture creation studies that have emerged since the early 1980s lack cohesiveness, and the domain remains controversial. This article not only exposes the limitations of the existing body of understanding on the topic but attempts to develop a more comprehensive and comprehendible framework for start up (new venture creation. To do so it uses the frameworks proposed by Whetten, and March and Smith to develop 11 propositions. The resultant model suggests that the creation of a start up involves the identification of an idea or opportunity by an entrepreneur who subsequently organizes a series of activities, mobilizes resources and creates competence using his/her networks in an environment in order to create value. It sheds light on the start-up (new venture creation process and has relevance for entrepreneurs, policy makers and researchers.

  13. Personality Traits and the Evaluation of Start-up Subsidies

    NARCIS (Netherlands)

    Caliendo, M.; Künn, Steffen; M., Weißenberger,

    2016-01-01

    Many countries support business start-ups to spur economic growth and reduce unemployment with different programmes. Evaluation studies of such programmes commonly rely on the conditional independence assumption (CIA), allowing a causal interpretation of the results only if all relevant variables

  14. A Boom Time for Education Start-Ups

    Science.gov (United States)

    DeSantis, Nick

    2012-01-01

    Harsh economic realities mean trouble for college leaders. But where administrators perceive an impending crisis, investors increasingly see opportunity. In recent years, venture capitalists have poured millions into education-technology start-ups, trying to cash in on a market they see as ripe for a digital makeover. And lately, those wagers have…

  15. Creating Start-up Companies around NCI Inventions | Poster

    Science.gov (United States)

    By Karen Surabian, Thomas Stackhouse, and Rose Freel, Contributing Writers, and Rosemarie Truman, Guest Writer The National Cancer Institute (NCI), led by the Technology Transfer Center (TTC),  the Avon Foundation, and The Center for Advancing Innovation have partnered to create a “first-of-a-kind” Breast Cancer Start-up Challenge.

  16. Start-Up Rhetoric in Eight Speeches of Barack Obama

    Science.gov (United States)

    O'Connell, Daniel C.; Kowal, Sabine; Sabin, Edward J.; Lamia, John F.; Dannevik, Margaret

    2010-01-01

    Our purpose in the following was to investigate the start-up rhetoric employed by U.S. President Barack Obama in his speeches. The initial 5 min from eight of his speeches from May to September of 2009 were selected for their variety of setting, audience, theme, and purpose. It was generally hypothesized that Barack Obama, widely recognized for…

  17. Dampak Mentoring Pada Keberhasilan Start-Up Business: Studi Kasus Pada Start-Up Business di Indonesia [Mentoring the Impact of Success of a Start-Up Business: A Case Study of a Start-Up Business in Indonesia

    Directory of Open Access Journals (Sweden)

    Christina Yanita Setyawati

    2016-10-01

    Full Text Available Entrepreneurship is a highly developed science today, as well as in the world of education in Indonesia. Teaching entrepreneurship at the university level, especially at some universities in Surabaya, the second largest city in Indonesia, shows the positive impact that there were 74.03% of start-up businesses that survived and thrived in 2014. Based on the previous observation, some business start-ups are accompanied by a mentor. Mentoring is a process of forming and maintaining lasting relationships that develop intensively among seniors with juniors. The mentoring function includes career and psychosocial functions. The main purpose of this research is to study the relationship between mentoring and the success of a start-up business. This study is a quantitative descriptive study using a random sampling method to obtain 100 samples. The tool used to collect data was a questionnaire. The methodology used to analyze the quantitative description to test the hypothesis. The results of this study indicate that mentoring influences the success of a start-up business that was owned by university students in Surabaya, Indonesia by 32% and 78% was influenced by other variables.

  18. Start-up of a drinking water biofilter

    DEFF Research Database (Denmark)

    Ramsay, Loren; Søborg, Ditte; Breda, Inês Lousinha Ribeiro

    When virgin filter media is placed in drinking water biofilters, a start-up period of some months typically ensues. During this period, the necessary inorganic coating and bacterial community are established on the filter medium, after which the treated water complies with drinking water criteria...

  19. Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie

    1998-01-01

    Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

  20. Synergies in the design and development of fusion and generation IV fission reactors

    International Nuclear Information System (INIS)

    Bogusch, E.; Carre, F.; Knebel, J.; Aoto, K.

    2007-01-01

    Future fusion reactors or systems and Generation IV fission reactors are designed and developed in worldwide programmes mostly involving the same partners to investigate and assess their potential for realisation and contribution to meet the future energy needs beyond 2030. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core face similar design issues and development needs. Therefore the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactors or systems will be designed for helium and liquid metal cooling and higher temperatures similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches might create synergistic design and development programmes. Therefore an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in support of common technologies. (orig.)

  1. Synergies in the design and development of fusion and generation IV fission reactors

    International Nuclear Information System (INIS)

    Bogusch, E.; Carre, F.; Knebel, J.U.; Aoto, K.

    2008-01-01

    Future fusion reactor and Generation IV fission reactor systems are designed and developed in worldwide programmes to investigate and assess their potential for realisation and contribution to the future energy needs beyond 2030 mostly involving the same partners. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except for the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core, face similar design issues and development needs. Therefore, the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactor systems will be designed for high-temperature helium and liquid metal cooling but also water including supercritical water and molten salt similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches can create synergistic design and development programmes. Therefore, an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in

  2. DIRECT ENERGY CONVERSION (DEC) FISSION REACTORS - A U.S. NERI PROJECT

    International Nuclear Information System (INIS)

    Beller, D.; Polansky, G.

    2000-01-01

    The direct conversion of the electrical energy of charged fission fragments was examined early in the nuclear reactor era, and the first theoretical treatment appeared in the literature in 1957. Most of the experiments conducted during the next ten years to investigate fission fragment direct energy conversion (DEC) were for understanding the nature and control of the charged particles. These experiments verified fundamental physics and identified a number of specific problem areas, but also demonstrated a number of technical challenges that limited DEC performance. Because DEC was insufficient for practical applications, by the late 1960s most R and D ceased in the US. Sporadic interest in the concept appears in the literature until this day, but there have been no recent programs to develop the technology. This has changed with the Nuclear Energy Research Initiative that was funded by the U.S. Congress in 1999. Most of the previous concepts were based on a fission electric cell known as a triode, where a central cathode is coated with a thin layer of nuclear fuel. A fission fragment that leaves the cathode with high kinetic energy and a large positive charge is decelerated as it approaches the anode by a charge differential of several million volts, it then deposits its charge in the anode after its kinetic energy is exhausted. Large numbers of low energy electrons leave the cathode with each fission fragment; they are suppressed by negatively biased on grid wires or by magnetic fields. Other concepts include magnetic collimators and quasi-direct magnetohydrodynamic generation (steady flow or pulsed). We present the basic principles of DEC fission reactors, review the previous research, discuss problem areas in detail and identify technological developments of the last 30 years relevant to overcoming these obstacles. A prognosis for future development of direct energy conversion fission reactors will be presented

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

    International Nuclear Information System (INIS)

    Osipov, S.L.; Tsikunov, A.G.; Lisitsin, E.C.

    1996-01-01

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

  4. Markets for reactor-produced non-fission radioisotopes

    International Nuclear Information System (INIS)

    Bennett, R.G.

    1995-01-01

    Current market segments for reactor produced radioisotopes are developed and reported from a review of current literature. Specific radioisotopes studied in is report are the primarily selected from those with major medical or industrial markets, or those expected to have strongly emerging markets. Relative market sizes are indicated. Special emphasis is given to those radioisotopes that are best matched to production in high flux reactors such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. A general bibliography of medical and industrial radioisotope applications, trends, and historical notes is included

  5. Investigations of the natural fission reactor program. Progress report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Cowan, G.A.; Norris, A.E.

    1978-10-01

    The U.S. study of the Oklo natural reactor began in 1973 with the principal objectives of understanding the processes that produced the reactor and that led to the retention of many of its products. Major facets of the program have been the chemical separation and mass spectrometric analysis of the reactor components and products, the petrological and mineralogical examination of samples taken from the reactor zones, and an interdisciplinary modeling of possible processes consistent with reactor physics, geophysics, and geochemistry. Most of the past work has been on samples taken within the reactor zones. Presently, these studies give greater emphasis to the measurement of mobile products in additional suites of samples collected peripherally and ''downstream'' from the reactor zones. This report summarizes the current status of research and the views of U.S. investigators, with particular reference to the extensive work of the French scientists, concerning the main features of the Oklo natural fission reactor. Also mentioned briefly is the U.S. search for natural fission reactors at other locations

  6. Thermodynamics of soluble fission products cesium and iodine in the Molten Salt Reactor

    NARCIS (Netherlands)

    Capelli, E.; Beneš, O.; Konings, R.J.M.

    2018-01-01

    The present study describes the full thermodynamic assessment of the Li,Cs,Th//F,I system. The existing database for the relevant fluoride salts considered as fuel for the Molten Salt Reactor (MSR) has been extended with two key fission products, cesium and iodine. A complete evaluation of all

  7. A method for measuring power signal background and source strength in a fission reactor

    International Nuclear Information System (INIS)

    Baers, B.; Kall, L.; Visuri, P.

    1977-01-01

    Theory and experimental verification of a novel method for measuring power signal bias and source strength in a fission reactor are reported. A minicomputer was applied in the measurements. The method is an extension of the inverse kinetics method presented by Mogilner et al. (Auth.)

  8. Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Zhang-cheng; Xie, Heng

    2014-01-01

    The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

  9. Thermochemical data for reactor materials and fission products: The ECN database

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1993-02-01

    The activities of the authors regarding the compilation of a database of thermochemical properties for reactor materials and fission products is reviewed. The evaluation procedures and techniques are outlined and examples are given. In addition, examples of the use of thermochemical data for the application in the field of Nuclear Technology are given. (orig.)

  10. Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Jian Hongbing

    1992-01-01

    The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

  11. Assessment of the high temperature fission chamber technology for the French fast reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; Geslot, B.; Domenech, T.; Normand, S. [Commissariat a l' Energie Atomique, CEA (France)

    2011-07-01

    High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

  12. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    Energy Technology Data Exchange (ETDEWEB)

    Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-05-15

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  13. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    International Nuclear Information System (INIS)

    Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-01-01

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  14. The role of relationships in start-up development

    DEFF Research Database (Denmark)

    Mattsson, Jan; Helmersson, Helge; Standing, Craig

    2017-01-01

    Relationships are important in the business start-up phase for a variety of reasons. Internal relationships can support knowledge exchange that determines the business model development and external relationships can facilitate a wide range of opportunities, support and insights. The paper explains...... the key relationships experienced by the founding team of a recently formed Swedish digital trading platform. Data were gathered through a self-reporting diary approach based on the Critical Incident Technique format and texts were analysed by the Pertex text-analytic software. The findings explain how...... important relationships were formed during the initial start-up period targeting international expansion. Interaction with early adopters enabled a rapid evolution of the business platform with the aim to build a community of users to support development....

  15. Model citizens. Outsourcing helps start-up Medicare HMO.

    Science.gov (United States)

    Slavic, B; Adami, S

    1999-04-01

    Health Plans of Pennsylvania (HPP), the managed care arm of Crozer-Keystone Health System, in Media, Pa. Selecting the information systems and building the infrastructure to support the start-up of a new Medicare HMO product. HPP chose to outsource the information systems needed to integrate all the components of managed care administration into a cost-effective and cohesive program. Because of its aggressive programming and start-up of the MedCarePlus product offering, HPP became the first plan in the country to submit Medicare claims data electronically for encounter reporting to the Health Care Financing Administration (HCFA). "Through an integrated team approach, an organization truly can benefit from the economies of scale gained through outsourcing."

  16. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2014-01-01

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM

  17. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tariq Siddique, M.; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-05-15

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM.

  18. The Prevessin Control Room during LEP's start up in 1989.

    CERN Multimedia

    1989-01-01

    The Prévessin Control Room saw its first momentous event when the 400 GeV beam for the SPS was commissioned in the presence of Project Leader John Adams. It was also here that the first proton-antiproton collisions were observed, in 1981. Eight years later, in 1989, operators and directors alike jumped for joy at the announcement of the first electron-positron collisions at the start up of LEP, the biggest accelerator in the world.

  19. ECommerce Strategy for Footwear Manufacturing Start-up

    OpenAIRE

    Kuzmanovic, Slobodan

    2010-01-01

    This paper defines and analyzes an eCommerce strategy for a footwear manufacturing start-up firm and assesses the financial impact this eCommerce strategy has in supporting the firm’s business model. The market analysis of the importance of eCommerce determined the online channel’s market share and influence on consumer purchasing decisions, while the industry analysis examined the competitive landscape and industry forces as drivers of forward integration from manufacturing to wholesale and ...

  20. A Start-Up Produces Infomercials for the Internet

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    @@ The one thing online shopping is missing, says Sukhinder Singh Cassidy, a former Google executive, is infomercials. Ms.Singh Cassidy introduced her new shopping start-up, Joyus.It is an e-commerce site that produces videos to sell products, like a Web version of the Home Shopping Network. "If you think about the ways to sell online, video is underutilized," she said."It's a new way for brands to merchandise their product."

  1. Digital Marketing practices amongst start-up accelerators

    OpenAIRE

    Azinheiro, Marisa Filipa Ramos

    2017-01-01

    Digital Marketing (DM) is a vital marketing tool used by all types of companies nowadays. Accelerators are companies that appeared in the last few years to help start-ups grow and, just like any other company, they are using DM as well. This research explored which were the DM strategies used by accelerators. To do so, an online survey was shared amongst more than 300 accelerators across the world, whose results were analyzed by using SPSS Software. Correlation analyses and significance te...

  2. Business start-ups and the effect of coaching programs

    OpenAIRE

    Loersch, Christian

    2015-01-01

    Entrepreneurship is known to be a main driver of economic growth. Hence, governments have an interest in supporting and promoting entrepreneurial activities. Start-up subsidies, which have been analyzed extensively, only aim at mitigating the lack of financial capital. However, some entrepreneurs also lack in human, social, and managerial capital. One way to address these shortcomings is by subsidizing coaching programs for entrepreneurs. However, theoretical and empirical evidence about busi...

  3. Opportunity identification and evaluation in franchisee business start-ups

    OpenAIRE

    Brookes, Maureen; Altinay, Levent; Wang, Xuan Lorna; Yeung, Ruth

    2016-01-01

    Purpose - This paper examines franchisees’ business start-ups from an entrepreneurial perspective, adopting a process representative of entrepreneurship to examine opportunity identification and evaluation by franchisees and to analyse factors that influence this process.\\ud \\ud Design/methodology/approach - A qualitative study was employed and data collected using semi-structured interviews with a sample of service industry franchisees in Macau.\\ud \\ud Findings - The study identifies that so...

  4. Anticipated simulation of Angra-1 start-up physical tests

    International Nuclear Information System (INIS)

    Fernandes, V.B.; Perrotta, J.A.; Silva Ipojuca, T. da; Ponzoni Filho, P.

    1981-01-01

    Some results foreseen by the Department of Nuclear Fuel (DCN. O) in Furnas for the measurements that will be realized during the start-up integrated tests for Angra-1 are presented. All the forecasting is based on a DCN.O proper correlation methodology, developed from basic physical principles, using computer codes developed by the authors or public computer codes adapted to this methodology. (E.G.) [pt

  5. Long Pulse EBW Start-up Experiments in MAST

    Directory of Open Access Journals (Sweden)

    Shevchenko V.F.

    2015-01-01

    Full Text Available Start-up technique reported here relies on a double mode conversion (MC for electron Bernstein wave (EBW excitation. It consists of MC of the ordinary (O mode, entering the plasma from the low field side of the tokamak, into the extraordinary (X mode at a mirror-polarizer located at the high field side. The X mode propagates back to the plasma, passes through electron cyclotron resonance (ECR and experiences a subsequent X to EBW MC near the upper hybrid resonance (UHR. Finally the excited EBW mode is totally absorbed at the Doppler shifted ECR. The absorption of EBW remains high even in cold rarefied plasmas. Furthermore, EBW can generate significant plasma current giving the prospect of a fully solenoid-free plasma start-up. First experiments using this scheme were carried out on MAST [1]. Plasma currents up to 33 kA have been achieved using 28 GHz 100kW 90ms RF pulses. Recently experimental results were extended to longer RF pulses showing further increase of plasma currents generated by RF power alone. A record current of 73kA has been achieved with 450ms RF pulse of similar power. The current drive enhancement was mainly achieved due to RF pulse extension and further optimisation of the start-up scenario.

  6. Actions needed for RA reactor exploitation - I-IV, Part I, Thermotechnical experiments related to RA reactor hot start-up; Radovi za potrebe eksploatacije reaktora RA - I-IV, I Deo, Termotehnicki eksperimenti u vezi pustanja rektora RA na snagu

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Heavy water coolant loop of the RA reactor includes the reactor, circulation pumps, heat exchangers and pipes. The objective of this task was measuring the thermal parameters of the RA reactor during operation. This report contains the results of the experiment, calculations of thermal regime for the outer and inner tubes, maximum temperature of the fuel element, fluid flow rate in the reactor channels, temperature of the coolant and fuel element cladding.

  7. The temperature distribution in a gas core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C. (Interuniversitair Reactor Inst., Delft (Netherlands)); Kistemaker, J.; Boersma-Klein, W.; Vitalis, F. (FOM-Instituut voor Atoom-en Molecuulfysica, Amsterdam (Netherlands))

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author).

  8. The temperature distribution in a gas core fission reactor

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C.; Kistemaker, J.; Boersma-Klein, W.; Vitalis, F.

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author)

  9. Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

    International Nuclear Information System (INIS)

    Wang Xinhua; Guo Haiping; Mou Yunfeng; Zheng Pu; Liu Rong; Yang Xiaofei; Yang Jian

    2013-01-01

    A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D + beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the three-dimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S(α, β) thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors. (authors)

  10. Fusion-fission hybrid as an alternative to the fast breeder reactor

    International Nuclear Information System (INIS)

    Barrett, R.J.; Hardie, R.W.

    1980-09-01

    This report compares the fusion-fission hybrid on the plutonium cycle with the classical fast breeder reactor (FBR) cycle as a long-term nuclear energy source. For the purpose of comparison, the current light-water reactor once-through (LWR-OT) cycle was also analyzed. The methods and models used in this study were developed for use in a comparative analysis of conventional nuclear fuel cycles. Assessment areas considered in this study include economics, energy balance, proliferation resistance, technological status, public safety, and commercial viability. In every case the characteristics of all fuel cycle facilities were accounted for, rather than just those of the reactor

  11. High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products

    International Nuclear Information System (INIS)

    Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

    1993-01-01

    An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ''engineered safety features,'' which, along with the use of high temperature capable materials further enhance its safety characteristics

  12. In-reactor testing of self-powered neutron detectors and miniature fission chambers

    International Nuclear Information System (INIS)

    Duchene, J.; LeMeur, R.; Verdant, R.

    1975-01-01

    The CEA has tested a variety of ''slow'' self-powered neutron detectors with rhodium, silver and vanadium emitters. Currently there are 120 vanadium detectors in the EL4 heavy water reactor. In addition, ''fast'' detectors with cobalt emitters have been tested at Saclay and 50 of these are in reactor. Other studies are concerned with 6 mm diameter miniature fission chambers. Two fast response chambers with argon-nitrogen filling gas became slow during irradiation, but operated to 600 deg C. An argon filled chamber of 4.7 mm diameter, for traversing in core system in pressurized water reactor, has shown satisfactory test results. (author)

  13. Transfer parameters of fission and activation products present in effluents of nuclear power reactors

    International Nuclear Information System (INIS)

    Cancio, D.; Menossi, C.A.; Ciallella, N.R.

    1978-01-01

    The paper presents results of research carried out in Argentina on transfer parameters of fission and activation products which may be present in the effluents of nuclear power reactors. For some nuclides, as Sr-90, Co-137 and I-131, the parameters were obtained by studies of the fallout, from measurements of integrated levels in the environment and in the food chains. Other values are concentration factors derived from laboratory and field experiments. They refer to fish, molluscs, crustaces and fresh water plants, for several fission and activation nuclides. Transfer parameters obtained have been of significant importance for environmental assessments, relating to nuclear installations in Argentina. (author)

  14. Transmutation of fission products in reactors and accelerator-driven systems

    International Nuclear Information System (INIS)

    Janssen, A.J.

    1994-01-01

    Energy flows and mass flows in several scenarios are considered. Economical and safety aspects of the transmutation scenarios are compared. It is difficult to find a sound motivation for the transmutation of fission products with accelerator-driven systems. If there would be any hesitation in transmuting fission products in nuclear reactors, there would be an even stronger hesitation to use accelerator-driven systems, mainly because of their lower energy efficiency and their poor cost effectiveness. The use of accelerator-driven systems could become a 'meaningful' option only if nuclear energy would be banished completely. (orig./HP)

  15. Inventories of radioactive fission products in the core of thermal nuclear reactor

    International Nuclear Information System (INIS)

    Marinkovic, N.

    1977-01-01

    As a part of the analysis concerning radiological consequences of a major LWR accident, inventories of the most significant radioactive nuclides and stable fission gases in the core of a PWR type reactor have been calculated. Calculations were performed by the DELFIN code using nuclide data and neutron flux data earlier obtained by the METHUSELAH code. Comparison with simplified calculation method show that it is quite rough for certain nuclides but the accuracy may be sufficient for safety analysis purposes recalling the inaccuracies in the later parts of fission product transport process (author)

  16. An assessment of fission product data for decay power calculation in fast reactors

    International Nuclear Information System (INIS)

    Sridharan, M.S.; Murthy, K.P.N.

    1987-01-01

    A review of our present capability at IGC, Kalpakkam to predict fission product decay power in fast reactors is presented. This is accomplished by comparing our summation calculations with the calculations of others and the reported experimental measurements. Our calculations are based on Chandy code developed at our Centre. The fission product data base of Chandy is essentially drawn from the yield data compiled by Crouch (1977) and the data on halflives etc. compiled by Tobias (1973). In general, we find good agreement amongst the different calculations (within ±5%) and our calculations also compare well with experimental measurements of AKIAMA et al and MURPHY et al

  17. Total Absorption Spectroscopy of Fission Fragments Relevant for Reactor Antineutrino Spectra and Decay Heat Calculations

    Directory of Open Access Journals (Sweden)

    Porta A.

    2016-01-01

    Full Text Available Beta decay of fission products is at the origin of decay heat and antineutrino emission in nuclear reactors. Decay heat represents about 7% of the reactor power during operation and strongly impacts reactor safety. Reactor antineutrino detection is used in several fundamental neutrino physics experiments and it can also be used for reactor monitoring and non-proliferation purposes. 92,93Rb are two fission products of importance in reactor antineutrino spectra and decay heat, but their β-decay properties are not well known. New measurements of 92,93Rb β-decay properties have been performed at the IGISOL facility (Jyväskylä, Finland using Total Absorption Spectroscopy (TAS. TAS is complementary to techniques based on Germanium detectors. It implies the use of a calorimeter to measure the total gamma intensity de-exciting each level in the daughter nucleus providing a direct measurement of the beta feeding. In these proceedings we present preliminary results for 93Rb, our measured beta feedings for 92Rb and we show the impact of these results on reactor antineutrino spectra and decay heat calculations.

  18. Start-up and ramp-up of the PLT tokamak by lower hybrid waves

    International Nuclear Information System (INIS)

    Jobes, F.C.; von Goeler, S.; Bernabei, S.

    1985-01-01

    Lower hybrid current drive (LHCD) is an inherently steady-state means of maintaining the poloidal field of a tokamak reactor. However, the energy losses of LHCD, which are proportional to density, are projected to be too great in a fusion reactor for LHCD to be economically feasible during the burn state of the reaction cycle. The authors maintain that LHCD could be extremely useful in restoring poloidal field energy between burns. In situations not requiring a rapid build up, LHCD appears, by extrapolation from present experiments, to be capable of supplying the full required poloidal field energy. In this paper, experiments have been performed on PLT and other tokamaks to examine the role of LHCD in start-up and ramp-up, as well as to examine the efficiency of stady-state current drice. Both the start-up and the ramp-up experiments were quite successful, with the start-up experiment obtaining currents up to 20% of full current for PLT, and the ramp-up experiments obtaining ramp-up efficiencies of approximately 20%

  19. Bootstrap finance: the art of start-ups.

    Science.gov (United States)

    Bhide, A

    1992-01-01

    Entrepreneurship is more popular than ever: courses are full, policymakers emphasize new ventures, managers yearn to go off on their own. Would-be founders often misplace their energies, however. Believing in a "big money" model of entrepreneurship, they spend a lot of time trying to attract investors instead of using wits and hustle to get their ideas off the ground. A study of 100 of the 1989 Inc. "500" list of fastest growing U.S. start-ups attests to the value of bootstrapping. In fact, what it takes to start a business often conflicts with what venture capitalists require. Investors prefer solid plans, well-defined markets, and track records. Entrepreneurs are heavy on energy and enthusiasm but may be short on credentials. They thrive in rapidly changing environments where uncertain prospects may scare off established companies. Rolling with the punches is often more important than formal plans. Striving to adhere to investors' criteria can diminish the flexibility--the try-it, fix-it approach--an entrepreneur needs to make a new venture work. Seven principles are basic for successful start-ups: get operational fast; look for quick break-even, cash-generating projects; offer high-value products or services that can sustain direct personal selling; don't try to hire the crack team; keep growth in check; focus on cash; and cultivate banks early. Growth and change are the start-up's natural environment. But change is also the reward for success: just as ventures grow, their founders usually have to take a fresh look at everything again: roles, organization, even the very policies that got the business up and running.

  20. Start-Up of FEL Oscillator from Shot Noise

    International Nuclear Information System (INIS)

    Kumar, V.; Krishnagopal, S.; Fawley, W.M.

    2007-01-01

    In free-electron laser (FEL) oscillators, as in self-amplified spontaneous emission (SASE) FELs, the buildup of cavity power starts from shot noise resulting from the discreteness of electronic charge. It is important to do the start-up analysis for the build-up of cavity power in order to fix the macropulse width from the electron accelerator such that the system reaches saturation. In this paper, we use the time-dependent simulation code GINGER [1]to perform this analysis. We present results of this analysis for the parameters of the Compact Ultrafast TErahertz FEL (CUTE-FEL) [2] being built at RRCAT

  1. Characteristics Contributing to High-Technology Start-Up Success

    DEFF Research Database (Denmark)

    Goslin, L.; Brown, W.; Palm, T.

    1993-01-01

    that the professional venture-capital-funded start-up firm will generally reflect the practices of perceptive management regarding: style, team approach, structure of the firm, market definition, culture, vision, and a formal strategy. Nevertheless, there are some questions about differences those characteristics...... by interview to provide insights on: the entrepreneurial-marketing style utilized, cultural characteristics of the firm, and perspectives of the time horizon of strategy. The representative firms (seven firms to a set) from whom information was solicited were drawn from the U.S. Pacific Northwest), Denmark...

  2. Research Strategies in Science-based Start-ups

    DEFF Research Database (Denmark)

    Valentin, Finn; Dahlgren, Johan Henrich; Lund Jensen, Rasmus

    develop a contingency view on complex problem solving which structures the argument into three steps:1) Characterising the problem architectures addressed by different types of DBFs;2) Testing and confirming that DBFs form requisite research strategies, by which we refer to problem solving approaches......Although biotech start-ups fail or succeed based on their research few attempts have been made to examine if and how they strategize in this core of their activity. Popular views on Dedicated Biotech Firms (DBFs) see the inherent uncertainty of research as defying notions of strategizing, directing...

  3. Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

    International Nuclear Information System (INIS)

    Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

    2006-01-01

    A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

  4. Fission track dating method: I. Study of neutron flux uniformity in some irradiation positions of IEA-R1 reactor

    International Nuclear Information System (INIS)

    Osorio, A.M.; Hadler, J.C.; Iunes, P.J.; Paulo, S.R. de

    1993-06-01

    In order to use the fission track dating method the flux gradient was verified within the sample holder, in some irradiation positions of the IEA-R1 reactor at IPEN/CNEN, Sao Paulo. The fission track dating method considers only the thermal neutron fission tracks, to subtract the other contributions sample irradiations with a cadmium cover was performed. The neutron flux cadmium influence was studied. (author)

  5. Journey from discovery of nuclear fission to accelerator-driven sub-critical reactor systems (ADS)

    International Nuclear Information System (INIS)

    Kapoor, S.S.

    2005-01-01

    The epoch making discovery of nuclear fission in 1939, which resulted purely from the curiosity driven basic research to understand the atomic and nuclear structure has changed the world forever with the onset of a new era in the history of human civilization. The basic nuclear physics research pursued after the discovery of fission has also been of much relevance in the harnessing of nuclear energy. In the recent years, there is considerable interest towards developing accelerator driven sub-critical reactor systems (ADS) for the incineration of the long-lived spent fuel radioactive waste and for the utilization of thorium fuel for nuclear power generation. In this talk, we discuss important milestones in the journey from discovery of nuclear fission to ADS. (author)

  6. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behaviour of simulant fission product species such as caesium iodide, caesium hydroxide and tellurium, in terms of their vapour deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high-density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO/sub 2/ clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapour phase, and specific data using this technique are reported

  7. Chemical aspects of fission product transport in the primary circuit of a light water reactor

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Dickinson, S.; Nichols, A.L.; Ogden, J.S.; Potter, P.E.

    1985-01-01

    The transport and deposition of fission products in the primary circuit of a light water reactor are of fundamental importance in assessing the consequences of severe accidents. Recent experimental studies have concentrated upon the behavior of simulant fission product species such as cesium iodide, cesium hydroxide and tellurium, in terms of their vapor deposition characteristics onto metals representative of primary circuit materials. An induction furnace has been used to generate high density/structural materials aerosols for subsequent analysis, and similar equipment has been incorporated into a glove-box to study lightly-irradiated UO 2 clad in Zircaloy. Analytical techniques are being developed to assist in the identification of fission product chemical species released from the fuel at temperatures from 1000 to 2500 0 C. Matrix isolation-infrared spectroscopy has been used to identify species in the vapor phase, and specific data using this technique are reported

  8. Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Corse, B.J.; Thompson, W.T.; Kaye, M.H.; Iglesias, F.C.; Elder, P.; Dickson, R.; Liu, Z.

    1997-01-01

    An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

  9. Start-up of horizontal anaerobic reactors with sludge blanket and fixed bed for wastewater treatment from coffee processing by wet method Partida de reatores anaeróbios horizontais com manta de lodo e de leito fixo para tratamento de águas residuárias do beneficiamento de frutos do cafeeiro por via úmida

    Directory of Open Access Journals (Sweden)

    Roberto A. de Oliveira

    2013-04-01

    Full Text Available In this study it was evaluated the start-up procedures of anaerobic treatment system with three horizontal anaerobic reactors (R1, R2 and R3, installed in series, with volume of 1.2 L each. R1 had sludge blanket, and R2 and R3 had half supporter of bamboo and coconut fiber, respectively. As an affluent, it was synthesized wastewater from mechanical pulping of the coffee fruit by wet method, with a mean value of total chemical oxygen demand (CODtotal of 16,003 mg L-1. The hydraulic retention time (HRT in each reactor was 30 h. The volumetric organic loading (VOL applied in R1 varied from 8.9 to 25.0 g of CODtotal (L d-1. The mean removal efficiencies of CODtotal varied from 43 to 97% in the treatment system (R1+R2+R3, stabilizing above 80% after 30 days of operation. The mean content of methane in the biogas were of 70 to 76%, the mean volumetric production was 1.7 L CH4 (L reactor d-1 in the system, and the higher conversions were around at 0.20 L CH4 (g CODremoved-1 in R1 and R2. The mean values of pH in the effluents ranged from 6.8 to 8.3 and the mean values of total volatile acids remained below 200 mg L-1 in the effluent of R3. The concentrations of total phenols of the affluent ranged from 45 to 278 mg L-1, and the mean removal efficiency was of 52%. The start-up of the anaerobic treatment system occurred after 30 days of operation as a result of inoculation with anaerobic sludge with active microbiota.Foram avaliados os procedimentos de partida de sistema de tratamento com três reatores anaeróbios horizontais (R1, R2 e R3, instalados em série, com volume de 1,2 L cada. O R1 com manta de lodo e o R2 e R3 através de suporte de bambu e fibra de coco, respectivamente. Como afluente,foram sintetizadas águas residuárias do despolpamento mecânico dos frutos do cafeeiro por via úmida, com valor médio de demanda química de oxigênio total (DQOtotal de 16.003 mg L-1. O tempo de detenção hidráulica (TDH em cada reator foi de 30 h. As

  10. Start-up First Term Sheet: Stumbling Blocks to Avoid

    Directory of Open Access Journals (Sweden)

    Amalyan Nataly D.

    2018-01-01

    Full Text Available Start-up relationships with business angel and/or venture capital are similar to a marriage – but with a divorce in mind. The latter necessitates a proper prenuptial agreement signing at the very beginning of the relations. In case of business angel and/or venture capital we are speaking about Term sheet, determining pricing, powers and duties of the board, as well as information, participation and protective rights for the partners. Standard Term sheet consists of three buckets of provisions, defining: (i terms impacting start-up valuation and economic division of profits and proceeds upon a liquidity event, (ii terms impacting control over decision making and (iii investor protection terms. This article focuses on financial and economic aspects of funding agreement, exposing some tricky parts of the Term sheet, threatening to transform potential win-win deals into rupture of relations, crash of the company or even legal battles. The results of the analysis are used to substantiate growing in popular appeal thesis that raising too much money can be detrimental for a startup.

  11. Qualitative and quantitative characteristics of fission products in spent nuclear fuel from RBMK-type reactor

    International Nuclear Information System (INIS)

    Adlys, G.; Adliene, D.

    2002-01-01

    Well-known empirical models or experimental instruments and methods for the estimation of fission product yields do not allow prediction of the behavior and evaluation of the time-dependent qualitative and quantitative characteristics of all fission products in spent nuclear fuel during long-term storage. Several computer codes were developed in different countries to solve this problem. French codes APOLLO1 and PEPIN were used in this work for modeling the characteristics of spent nuclear fuel in the RBMK reactor. The modeling results of qualitative and quantitative characteristics of long-lived fission products for different cooling periods of spent nuclear fuel, including 50-year cooling period, are presented in this paper. The 50-year cooling period conforms to the foreseen time of storage of spent nuclear fuel in CONSTOR and CASTOR casks at the Ignalina NPP. These results correlate well with evaluated quantities for the well-known yields of the nuclides and could be used for the compilation of the database for long-lived fission products in spent nuclear fuel from the RBMK-type reactor. They allow one to predict and to solve effectively safety problems concerning with long-term spent nuclear fuel storage in casks. (author)

  12. Geochemical properties and nuclear chemical characteristics of Oklo natural fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hidaka, Hiroshi [Hiroshima Univ., Higashi-Hiroshima (Japan). Faculty of Science

    1997-07-01

    There are six uranium deposits in the Gabonese Republic in the cnetral Africa. `Fission reactor zone`, the fission chain reactions generated about 200 billion years ago, was existed in a part of them. CEA begun geochemical researches of Oklo deposits etc. in 1991. The geochemical and nuclear chemical properties of Oklo were reviewed from the results of researches. Oklo deposits is consisted of main five sedimentary faces such as sandstone (FA), Black Shale formation (FB), mudstone (FC), tuff (FD) and volcaniclastic sandstone (FE) from the bottom on the base rock of granite in the Precambrian era. Uranium is enriched in the upper part of FA layer and the under part of FB layer. {sup 235}U/{sup 238}U, U content, fission proportion, duration time, neutron fluence, temperature, restitution factor of {sup 235}U and epithermal index ({gamma}) were investigated and compared. The geochemical properties of Oklo are as followed: large enrich of uranium, the abundance ratio of {sup 235}U as same as that of enriched uranium, interaction of natural water and small rear earth elements. These factors made casually Oklo fission reactor. (S.Y.)

  13. Molten fluoride mixtures as possible fission reactor fuels

    International Nuclear Information System (INIS)

    Grimes, W.R.

    1978-01-01

    Molten mixtures of fluorides with UF 4 as a component have been used as combined fuel and primary heat transfer agent in experimental high-temperature reactors and have been proposed for use in breeders or converters of 233 U from thorium. Such use places stringent and diverse demands upon the fluid fuel. A brief review of chemical behavior of molten fluorides is given to show some of their strengths and weaknesses for such service

  14. Application of Campbell's MSV method in monitoring of reactor's fission power

    International Nuclear Information System (INIS)

    Stankovic, S.J.; Vukcevic, M.; Loncar, B.; Vasic, A.; Osmokrovic, P.

    2003-01-01

    This paper presents some possibilities of Campbell's MSV (Mean Square Value) method in monitoring the reactor's fission power. Investigation of gamma discrimination compared to neutron component of signal along with change of variance and mean value the detector output signal for a specified range of reactor's fission power (10mW-22W) was carried out. The uncompensated ionization chamber for mixed n- gamma fields was used as detector element. Experimental measurements were performed using digitized MSV method, and obtained results were compared to those obtained by classical measuring chain. The final conclusion is that the order of discrimination in MSV signal processing is about fifty times larger than for classical measuring method (author)

  15. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    Science.gov (United States)

    Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

    2016-02-01

    The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  16. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    Directory of Open Access Journals (Sweden)

    Wagemans Jan

    2016-01-01

    Full Text Available The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  17. Conceptual design of a fission-based integrated test facility for fusion reactor components

    International Nuclear Information System (INIS)

    Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

    1982-01-01

    The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

  18. Major features of a mirror fusion--fast fission hybrid reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Burleigh, R.J.

    1974-01-01

    A conceptual design was made of a fusion-fission reactor. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and sustained by hot neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and is cooled by helium. It was shown how the reactor can be built using essentially present day construction technology and how the uranium bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel of which approximately 1200 kg of plutonium are produced each year along with the approximately 750 MW of electricity. (U.S.)

  19. Feasibility study of a fission supressed blanket for a tandem-mirror hybrid reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Barr, W.L.

    1981-01-01

    A study of fission suppressed blankets for the tandem mirror not only showed such blankets to be feasible but also to be safer than fissioning blankets. Such hybrids could produce enough fissile material to support up to 17 light water reactors of the same nuclear power rating. Beryllium was compared to 7 Li for neutron multiplication; both were considered feasible but the blanket with Li produced 20% less fissile fuel per unit of nuclear power in the reactor. The beryllium resource, while possibly being too small for extensive pure fusion application, would be adequate (with carefully planned industrial expansion) for the hybrid because of the large support ratio, and hence few hybrids required. Radiation damage and coatings for beryllium remain issues to be resolved by further study and experimentation. Molten salt reprocessing was compared to aqueous solution reprocessing

  20. Start-up of commercial high level waste vitrification facilities at La Hague

    International Nuclear Information System (INIS)

    Sombret, C.; Jouan, A.; Fournier, W.; Alexandre, D.; Leroy, L.

    1991-01-01

    The paper describes industial experience gained in France for vitrification of fission products generated by spent fuel reprocessing. The continuous vitrification process developed by CEA, SGN and COGEMA is outlined and Marcoule Vitrification Facility (AVM), with output results since start-up of hot operation in June 1978, briefly presented. Vitrification of high-level liquid waste has now entered an industrial phase at La Hague with R7 and T7 facilities. R7 and T7 have each been designed to process FP solutions generated by reprocessing LWR fuel with an initial enrichment of 3.5% and a discharge burn-up of 33,000 MWd/t. R7 active operations began on June, 1989. This facility is now vitrifying the backlog of fission products resulting from the existing UP2 reprocessing plant, which is being currently extended. Scheduled to start early in 1992, T7 will vitrify the fission products, dissolution fines and sodium-rich solutions issuing from UP3 plant

  1. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    International Nuclear Information System (INIS)

    Ellison, P.G.; Monson, P.R.; Mitchell, H.A.

    1990-01-01

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises

  2. Japanese list of requests for neutron nuclear data for fission reactors

    International Nuclear Information System (INIS)

    Igarasi, Sin-iti; Asami, Tetsuo

    1977-05-01

    Requests for neutron nuclear data for fission reactors are presented. These are screened by a WRENDA Working Group of Japanese Nuclear Data Committee and submitted to WRENDA 76/77. This report includes 163 requests of which 55 requests are newly registered in the WRENDA. Three requests of the previous list are withdrawn. This activity is a part of the international cooperation with CCDN, NEANDC and INDC. (auth.)

  3. Pressure due to fission gases in a fuel element circulating in a reactor

    International Nuclear Information System (INIS)

    Fonteray, Jean

    1965-01-01

    This document states calculation hypotheses and methods used to assess pressures due to fission gases in a fuel element moving in a reactor channel in the reverse direction with respect to the cooling fluid. The calculation comprises the calculation of the temperature in the fuel rod, of the reduced diffusion coefficient, of the diffused gas fraction, of the pressure. The appendix describes the use of the SPM 076 software: input data, output results, computing time [fr

  4. Feasibility study of a fission-suppressed tandem-mirror hybrid reactor

    International Nuclear Information System (INIS)

    Lee, J.D.; Moir, R.W.; Barr, W.L.

    1982-04-01

    Results of a conceptual design study of a U-233 producing fusion breeder consisting of a tandem mirror fusion device and two types of fission-suppressed blankets are presented. The majority of the study was devoted to the conceptual design and evaluation of the two blankets. However, studies in the areas of fusion engineering, reactor safety, fuel reprocessing, other fuel cycle issues, economics, and deployment were also performed

  5. Resuspension of fission products during severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    Borkowski, R.; Bunz, H.; Schoeck, W.

    1986-05-01

    This report investigates the influence of resuspension phenomena on the overall radiological source term of core melt accidents in a pressurized water reactor. A review of the existing literature is given and the literature data are applied to calculations of the source term. A large scatter in the existing data was found. Depending on the scenario and on the data set chosen for the calculations the relative influence of resuspended fission products on the source term ranges from dominant to negligible. (orig.) [de

  6. Thermodynamic performance of a gas-core fission reactor

    International Nuclear Information System (INIS)

    Klein, W.

    1987-01-01

    The purpose of this thesis was to investigate the thermodynamic behaviour of a critical quantity of gaseous uranium-fluorides in chemical equilibrium with a graphite wall. From the very beginning a container was considered with cooled walls. As it was evident that a nuclear reactor working with gaseous fuel should run at much higher temperatures than classical LWR or HTGR reactors, most of the investigations were performed for walls with a surface temperature of 1800 to 2000 K. It was supposed that such a surface temperature would be technologically possible for a heat load between 1 and 5 MWatt m -2 . Cooling with high pressure helium-gas has to keep balance with this heat flux. The technical construction of such a wall will be a problem in itself. It is thought that the experiences with re-entry-vessels in space-technology can be used. A basic assumption in all the calculations is that the U-C-F reactor gas 'sees' a graphite wall, possibly graphite tiles supported by heat resistant materials like SiN 2 , SiC 2 and at a lower temperature level by niobium-steel. Such a gastight compound-system is not necessarily of high-tensile strength materials. It has to be surrounded by a cooled neutron moderator-reflector which in its turn must be supported by a steel-wall at room temperature holding pressure of the order of 100 bar (10 MPa). The design of such a compound-wall is a task for the future. 116 refs.; 28 figs.; 29 tabs

  7. Simulation of fusion first-wall environment in a fission reactor

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Kulcinski, G.L.; Longhurst, G.R.

    1982-01-01

    A novel concept to produce a realistic simulation of a fusion first-wall test environment has been proposed recently. This concept takes advantage of the (/eta/, α) reaction in 59 Ni to produce a high internal helium content in the metal while using the 3 He (/eta/, /rho/)T reaction in the gas surrounding the specimen to produce an external heat and particle flux. Models to calculate heat flux, erosion rate, implantation, and damage rate to the walls of the test module are presented. Preliminary results show that a number of important fusion technology issues could be tested experimentally in a fission reactor such as the Engineering Test Reactor

  8. TFTR Mirnov coil analysis at plasma start-up

    International Nuclear Information System (INIS)

    Harley, T.R.; Buchenauer, D.A.; Coonrod, J.; McGuire, K.M.

    1986-01-01

    The methods for finding poloidal and toroidal numbers of MHD oscillations from Mirnov coils are reviewed and modified. Examples of various MHD phenomena occurring during start-up on TFTR are illustrated. It is found that the MHD mode structure best fits a model with the toroidal correction included. A new algorithm which finds m,n numbers can accommodate toroidal effects which are manifested in the phase data. The algorithm can find m,n numbers with a given toroidal correction parameter lambda', (lambda' = 0 → cylindrical). This algorithm is also used to find the optimal value of lambda' automatically, eliminating the need for ''guesswork.'' The algorithm finds the best parameters to the fit much faster than more conventional computational techniques. 9 refs., 21 figs., 2 tabs

  9. Conceptual design study of Hyb-WT as fusion–fission hybrid reactor for waste transmutation

    International Nuclear Information System (INIS)

    Siddique, Muhammad Tariq; Kim, Myung Hyun

    2014-01-01

    Highlights: • Conceptual design study of fusion-fission hybrid reactor for waste transmutation. • MCNPX and MONTEBURNS are compared for transmutation performance of WT-Hyb. • Detailed neutronic performance of final optimized Hyb-WT design is analyzed. • A new tube-in-duct core design is implemented and compared with pin type design. • Study shows many aspects of hybrid reactor even though scope was limited to neutronic analysis. - Abstract: This study proposes a conceptual design of a hybrid reactor for waste transmutation (Hyb-WT). The design of Hyb-WT is based on a low-power tokamak (less than 150 MWt) and an annular ring-shaped reactor core with metal fuel (TRU 60 w/o, Zr 40 w/o) and a fission product (FP) zone. The computational code systems MONTEBURNS and MCNPX2.6 are investigated for their suitability in evaluating the performance of Hyb-WT. The overall design performance of the proposed reactor is determined by considering pin-type and tube-in-duct core designs. The objective of such consideration is to explore the possibilities for enhanced transmutation with reduced wall loading from fusion neutrons and reduced transuranic (TRU) inventory. TRU and FP depletion is analyzed by calculating waste transmutation ratio, mass burned per full power year (in units of kg/fpy), and support ratio. The radio toxicity analysis of TRUs and FPs is performed by calculating the percentage of toxicity reduction in TRU and FP over a burn cycle

  10. Source driven breeding fission power reactors and the nuclear energy strategy

    International Nuclear Information System (INIS)

    Greenspan, E.

    The nuclear energy economy is facing severe difficulties associated with low utilization of uranium resources, safety, non-proliferation and environmental issues. Energy policy makers face the dilemma: commercialize LMFBRs immediately with the risk of negative economical, proliferation or other consequences, or continue with R and D programs that will provide the information needed for sounder decisions, but now taking the risk of running out of economically exploitable uranium ore resources. The development of hybrid reactors can provide an assurance against the latter risk and offers many interesting new options for the nuclear energy strategy. Being based on the technology of LWRs and HWRs, Light Water Hybrid Reactors (LWHR) provide a most natural link between the fission reactor technology of the present and the fusion power technology of the future. The investment in their development in excess of that required for the development of fusion power reactors is expected to be relatively small, thus making the development of LWHRs potentially a high benefit-to-cost ratio program. It is recommended that the fission and fusion communities will cooperate in hybrids R and D programs aimed at assessing the technological and economical viability of hybrid reactors as reliably and soon as possible. (author)

  11. Start-up phase assessment of a UASB-Septic tank system treating domestic septage

    International Nuclear Information System (INIS)

    Ali, M.; Al-Saed, R.; Mahmoud, N.

    2007-01-01

    About 65% of the annual domestic waste water in Palestine is currently collected in cesspits, where inadequate disposal might cause cumulative public health risks and annual environmental degradation. This research presents the preliminary results for the start-up period of a pilot-scale UASB-septic tank system treating domestic septage of Birzet town. Under different operational conditions, the performance of the pretreatment system for the removal of organic matter and nutrients was evaluated. Initial results showed that organic pollutants removal was mainly due to biophysical processes including sedimentation and microbial degradation. During start-up phase, the system attained removal efficiency for COD total of about 80% compared to removal for COD col, and COD dis of 71% and 43% respectively. Similarly, the continuous operation mode demonstrated that the system was quite effective in removing organic pollutants. Operational experience from the initial results revealed that seeding the USAB reactor with activated sludge during the start up period was not practical. Finally, the advantage of USAB-septic tank application appeared to be achievable if adequate system operation and control over a long monitoring period were maintained. (author)

  12. Careers at Biotech Start-Ups and in Entrepreneurship.

    Science.gov (United States)

    Froshauer, Susan

    2017-11-01

    The world of biotechnology "start-ups" and entrepreneurship offers exciting new avenues for driving state-of-the-art research using an arsenal of multidisciplinary skills, whether your role is as part of a team or as a leader. Although traditionally these positions may not be as secure as those offered by some of the larger companies, the small start-up culture provides opportunities for contributing at many levels to a wide range of responsibilities: from scientific discovery to delivery of proof of concept and intellectual property; from analysis of market opportunities and competitive intelligence to creation of time lines and business plans for a first product. Often, if you get in on the ground level, you get to validate your own concept, pitch to potential investors, argue value, build a team, engage advisors, and then, with funding in hand, launch an entirely new research and development (R&D) enterprise. Many of the skills and much of the experience gained while pursuing a graduate degree can be put to good use in these arenas as well. This path, however, is not for the faint of heart; it requires not only a strong scientific background and organizational skills, but also the ability to work well on a team, excellent communication skills, and persistence when faced with delays or disappointment. With increasing responsibilities in the small company come the requirements for aptitudes for leadership, strategic and financial planning, networking, negotiating, and managing both projects and personnel. Copyright © 2017 Cold Spring Harbor Laboratory Press; all rights reserved.

  13. Development of data processing system for the start-up test of FUGEN

    International Nuclear Information System (INIS)

    Nakajima, Ichiro; Kato, Hidemasa

    1981-01-01

    The data processing in the start-up test of conventional reactors has been carried out by recording data with transient phenomena recorders (e.g. electromagnetic oscillographs) or analog data recorders. On the other hand, the rapid works for detailed comparison and investigation between the test data and the analyzed results have been indispensable because ''Fugen'' is a new type of reactor, for which the results in conventional reactors did not necessarily serve as reference. Therefore, in the start-up test of the ''Fugen'' plant, the test data processing and the forecast analysis were performed by installing on the site a mini-computer capable of independently processing the test data and a terminal equipment connected to a large computer with a special communication line. As soon as the testing was completed, the comparison of the test data with the forecast analysis was presented on a graphic display (CRT), and the analysis was modified until significant differences did not appear between the test data and the analyzed data. In this paper, the system hardware and software are described, and two functions of forecast analysis and test data processing are explained. The time required for printing-out or graphic display from inputting 600 kB analysis code data using the terminal equipment was 10 to 30 minutes, and the evaluation and investigation for the test results were able to be immediately achieved by the data processing using the mini-computer. This is one of the factors to carry out the start-up test satisfactorily together with the forecast analysis works. (Wakatsuki, Y.)

  14. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  15. Start-up strategies for thermophilic anaerobic digestion of pig manure

    International Nuclear Information System (INIS)

    Moset, V.; Bertolini, E.; Cerisuelo, A.; Cambra, M.; Olmos, A.; Cambra-López, M.

    2014-01-01

    Sludge physicochemical composition, methane (CH 4 ) yield, and methanogenic community structure and dynamics using quantitative real-time polymerase chain reaction were determined after start-up of anaerobic digestion of pig manure. Eight thermophilic continuous stirred anaerobic digesters were used during 126 days. Four management strategies were investigated: a feedless and a non-feedless period followed by a gradual or an abrupt addition of pig manure (two digesters per strategy). During the first 43 days, VFA (volatile fatty acids) accumulations and low CH 4 yield were observed in all digesters. After this period, digesters recovered their initial status being propionic acid the last parameter to be re-established. Non-feedless digesters with an abrupt addition of pig manure showed the best performances (lower VFA accumulation and higher CH 4 yield). Differences in microbial orders and dynamics, however, were less evident among treatments. Hydrogenotrophic methanogenesis, Methanomicrobiales first and Methanobacteriales second, was the dominant metabolic pathway in all digesters. Further research is needed to clarify the role and activity of hydrogenotrophic methanogens during the recovery start-up period and to identify the best molecular tools and methodologies to monitor microbial populations and dynamics reliably and accurately in anaerobic digesters. - Highlights: • Four start-up strategies for thermophilic anaerobic digestion of pig manure were tested. • Physicochemical composition, methane yield and methanogenic community were determined. • During the first 43 days, a decline in reactor's performance occurred. • The best start-up strategy was non-feedless with an abrupt addition of pig slurry. • Hydrogenotrophic methanogenesis was the dominant metabolic pathway

  16. Study of advanced fission power reactor development for the United States. Volume II

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

  17. The Experiment Study of Anaerobic Ammonia Oxidation Start-up by Using the Upflow Double Layer Anaerobic Filter

    Directory of Open Access Journals (Sweden)

    YAO Li

    2018-02-01

    Full Text Available Anammox is an efficient nitrogen removal process, but it is difficult to start-up and operate, and ananammox reactor is the efficient way to resolve this problem. The start-up of anammox reactor by upflow anaerobic filter was studied. Denitrifying sludge, anaerobic sludge, and mixed sludge was inoculated on the packing materials, respectively and an autotrophic denitrification condition was provided by the simulated wastewater influent. Along with the gradual increase of matrix concentration and hydraulic load, the microflora was converted to the anaerobic ammonium oxidation(anammoxreaction. The results showed that the anammox reaction could be started by all the three sludge, and the time of start-up of denitrifying sludge, anaerobic sludge, mixed sludge was 42, 54 days and 45 days, respectively. The best result was that inoculated with denitrifying sludge with 82.2% of the total nitrogen removal rate, which started-up quickly and nitrogen was removed efficiently. Double packing effectively improved the stability of anammox process in the reactor, in which the suitable influent concentration loading for the anammox bacteria was 270 mg·L-1 and 360 mg·L-1 for ammonia nitrogen and nitrite nitrogen, respectively, and the COD concentration could not be more than 150 mg· L-1. Furthermore, there was a coexist-effect for anaerobic ammonia oxidation and methanation in this reactor system.

  18. Improving ADM1 model to simulate anaerobic digestion start-up with inhibition phase based on cattle slurry

    International Nuclear Information System (INIS)

    Normak, A.; Suurpere, J.; Suitso, I.; Jõgi, E.; Kokin, E.; Pitk, P.

    2015-01-01

    The Anaerobic Digestion Model No.1 (ADM1) was improved to simulate an anaerobic digestion start-up phase. To improve the ADM1, a combined hydrolysis equation was used based on the Contois model of bacterial growth and the function of hydrolysis inhibition by VFA. The start-up with fresh cattle slurry was carried out in a pilot-scale reactor to calibrate the chosen parameters of the ADM1. The important aspects of model calibration were hydrolysis rate, the number of anaerobic microbes in cattle slurry, and the growth rate of bacteria. Good simulation results were achieved after calibration for the independent start-up test with pre-conditioned cattle slurry. - Highlights: • Improved ADM1 can be used for simulation of reactor start-up with inhibition phase. • The hydrolysis rate had a decreased value in case of high VFA concentration or low number of hydrolytic bacteria. • Hydrolysis inhibitory threshold value of 9.85 g L −1 was obtained for VFA. • Start-up with pre-conditioned cattle slurry had a relatively short inhibition phase

  19. Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

    International Nuclear Information System (INIS)

    Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

    1978-01-01

    The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

  20. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs

  1. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs.

  2. Conceptual design of a hybrid fusion-fission reactor with intrinsic safety and optimized energy productivity

    International Nuclear Information System (INIS)

    Talebi, Hosein; Sadat Kiai, S.M.

    2017-01-01

    Highlights: • Designing a high yield and feasible Dense Plasma Focus for driving the reactor. • Presenting a structural method to design the dual layer cylindrical blankets. • Finding, the blanket production energy, in terms of its geometrical and material parameters. • Designing a subcritical blanket with optimization of energy amplification in detail. - Abstract: A hybrid fission-fusion reactor with a Dense Plasma Focus (DPF) as a fusion core and the dual layer fissionable blanket as the energy multiplier were conceptually designed. A cylindrical DPF, energized by a 200 kJ bank energy, is considered to produce fusion neutron, and these neutrons drive the subcritical fission in the surrounding blankets. The emphasis has been placed on the safety and energy production with considering technical and economical limitations. Therefore, the k eff-t of the dual cylindrical blanket was defined and mathematically, specified. By applying the safety criterion (k eff-t ≤ 0.95), the geometrical and material parameters of the blanket optimizing the energy amplification were obtained. Finally, MCNPX code has been used to determine the detailed dimensions of the blankets and fuel rods.

  3. Direct energy conversion in fission reactors: A U.S. NERI project

    International Nuclear Information System (INIS)

    Slutz, Stephen A.; Seidel, David B.; Polansky, Gary F.; Rochau, Gary E.; Lipinski, Ronald J.; Besenbruch, G.; Brown, L.C.; Parish, T.A.; Anghaie, S.; Beller, D.E.

    2000-01-01

    In principle, the energy released by a fission can be converted directly into electricity by using the charged fission fragments. The first theoretical treatment of direct energy conversion (DEC) appeared in the literature in 1957. Experiments were conducted over the next ten years, which identified a number of problem areas. Research declined by the late 1960's due to technical challenges that limited performance. Under the Nuclear Energy Research Initiative the authors are determining if these technical challenges can be overcome with todays technology. The authors present the basic principles of DEC reactors, review previous research, discuss problem areas in detail, and identify technological developments of the last 30 years that can overcome these obstacles. As an example, the fission electric cell must be insulated to avoid electrons crossing the cell. This insulation could be provided by a magnetic field as attempted in the early experiments. However, from work on magnetically insulated ion diodes they know how to significantly improve the field geometry. Finally, a prognosis for future development of DEC reactors will be presented

  4. Agile methodology selection criteria: IT start-up case study

    Science.gov (United States)

    Micic, Lj

    2017-05-01

    Project management in modern IT companies is often based on agile methodologies which have several advantages compared to traditional methodologies such is waterfall. Having in mind that clients sometimes change project during development it is crucial for an IT company to choose carefully which methodology is going to implement and is it going to be mostly based on one or is it going got be combination of several. There are several modern and often used methodologies but among those Scrum, Kanban and XP programming are usually the most common. Sometimes companies use mostly tools and procedures from one but quite often they use some of the combination of those methodologies. Having in mind that those methodologies are just a framework they allow companies to adapt it for their specific projects as well as for other limitations. These methodologies are in limited usage Bosnia but more and more IT companies are starting to use agile methodologies because it is practice and common not just for their clients abroad but also starting to be the only option in order to deliver quality product on time. However it is always challenging which methodology or combination of several companies should implement and how to connect it to its own project, organizational framework and HR management. This paper presents one case study based on local IT start up and delivers solution based on theoretical framework and practical limitations that case company has.

  5. Conceptual design of the IFMIF Start-Up monitoring module

    Energy Technology Data Exchange (ETDEWEB)

    Gouat, Philippe, E-mail: philippe.gouat@sckcen.be [SCK-CEN – The Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Leysen, Willem; Goussarov, Andrei; Galledou, Papa Sally [SCK-CEN – The Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Rapisarda, David; Mota, Fernando; Garcia, Angela [CIEMAT – Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, Avda. Complutense 40, 28040 Madrid (Spain)

    2013-10-15

    Highlights: ► IFMIF test module conceptual design. ► IFMIF test module foreseen instrumentation. ► Cerenkov photon flux monitor. -- Abstract: The preliminary engineering design of the test facilities, including the various test modules to be used in the IFMIF plant is a part of the IFMIF/EVEDA (Engineering Validation and Engineering Design Activities) project from the Broader Approach to fusion. One presents the current status of the conceptual development of the IFMIF Start-Up Monitoring Module, a dedicated device used in the IFMIF test cell during the commissioning phase of the installation, in order to completely characterise the irradiation conditions behind the target on which the beam of deuterons will be focused. This STUMM embarks a lot of instrumentation to precisely characterise the neutron field, the nuclear heating and the temperatures in the test cell. One briefly describes the measuring instruments (including a specific radiation flux monitor under development), the possible layouts and the possible positioning. One also defines which types of measurements are expected by this especially dedicated commissioning module.

  6. In-reactor measurements of thermo mechanical behaviour and fission gas release of water reactor fuel

    International Nuclear Information System (INIS)

    Kolstad, E.; Vitanza, C.

    1983-01-01

    the fuel performance during and after a power ramp can be investigated by direct in-pile measurements related to the thermal, mechanical and fission gas release behaviour. The thermal response is examined by thermocouples placed at the centre of the fuel. Such measurements allow the determination of thermal feedback effects induced by the simultaneous liberation of fission gases. The thermal feedback effect is also being separately studied out-of-pile in a specially designed rod where the fission gas release is simulated by injecting xenon in known quantities at different axial positions within the rod. Investigations on the mechanical behaviour are based on axial and diametral cladding deformation measurements. This enables the determination of the amount of local cladding strain and ridging during ramping, the extent of relaxation during the holding time and the amount of residual (plastic) deformation. Gap width measurements are also performed in operating fuel rods using a cladding deflection technique. Fission gas release data are obtained, besides from post-irradiation puncturing, by continuous measurements of the rod internal pressure. This type of measurement leads to the description of the kinetics of the fission gas release process at different powers. The data tend to indicate that the time-dependent release can be reasonably well described by simple diffusion. The paper describes measuring techniques developed and currently in use in Halden, and presents and discusses selected experimental results obtained during various power ramps and transients. (author)

  7. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  8. Decision Enhancement Services for Small and Medium Enterprise Start-Ups in Uganda

    NARCIS (Netherlands)

    Sol, H G; Basaza, Habinka; Sprague, Ralph H.

    2012-01-01

    While start-up firms create a substantial economic impact on most economies, the failure rate of start-up firms seems to remain high due to inadequate agile decision services. Deciding to start-up mining Small and Medium Enterprises (SME) is a challenging task in Uganda. Research on SME start-up

  9. Resonance self-shielding effect in uncertainty quantification of fission reactor neutronics parameters

    International Nuclear Information System (INIS)

    Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

    2014-01-01

    In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

  10. Survey on the fusion/fission-hybrid-reactors, a literature review

    International Nuclear Information System (INIS)

    A survey, based on existing literature, of the work being pursued worldwide on fusion - fission (hybrid) reactor systems is presented. Six areas are reviewed: Plasma physics parameters; Blankets concepts; Fuel cycles; Reactor conceptual designs; Safety and environmental problems; System studies and economic perspectives. Attention has been restricted to systems using magnetically confined plasmas, mainly to mirror and Tokamak - type concepts. The aim is to provide sufficient information, even if not exhaustive, on hybrid reactor concepts in order to help understand what may be expected from their possible development and the ways in which hybrids could affect the future energy scenario. Some concluding remarks are made which represent the personal view of the authors only

  11. Energy deposition measurements in fast reactor safety experiments with fission thermocouple detectors

    International Nuclear Information System (INIS)

    Wright, S.A.; Scott, H.L.

    1979-01-01

    The investigation of phenomena occurring in in-pile fast reactor safety experiments requires an accurate measurement of the time dependent energy depositions within the fissile material. At Sandia Laboratories thin-film fission thermocouples are being developed for this purpose. These detectors have high temperature capabilities (400 to 500 0 C), are sodium compatible, and have milli-second time response. A significant advantage of these detectors for use as energy deposition monitors is that they produce an output voltage which is directly dependent on the temperature of a small chip of fissile material within the detectors. However, heat losses within the detector make it necessary to correct the response of the detector to determine the energy deposition. A method of correcting the detector response which uses an inverse convolution procedure has been developed and successfully tested with experimental data obtained in the Sandia Pulse Reactor (SPR-II) and in the Annular Core Research Reactor

  12. RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS

    Directory of Open Access Journals (Sweden)

    GO CHIBA

    2014-06-01

    Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

  13. Needs and accuracy requirements for fission product nuclear data in the physics design of power reactor cores

    International Nuclear Information System (INIS)

    Rowlands, J.L.

    1978-01-01

    The fission product nuclear data accuracy requirements for fast and thermal reactor core performance predictions were reviewed by Tyror at the Bologna FPND Meeting. The status of the data was assessed at the Meeting and it was concluded that the requirements of thermal reactors were largely met, and the yield data requirements of fast reactors, but not the cross section requirements, were met. However, the World Request List for Nuclear Data (WRENDA) contains a number of requests for fission product capture cross sections in the energy range of interest for thermal reactors. Recent reports indicate that the fast reactor reactivity requirements might have been met by integral measurements made in zero power critical assemblies. However, there are requests for the differential cross sections of the individual isotopes to be determined in addition to the integral data requirements. The fast reactor requirements are reviewed, taking into account some more recent studies of the effects of fission products. The sodium void reactivity effect depends on the fission product cross sections in a different way to the fission product reactivity effect in a normal core. This requirement might call for different types of measurement. There is currently an interest in high burnup fuel cycles and alternative fuel cycles. These might require more accurate fission product data, data for individual isotopes and data for capture products. Recent calculations of the time dependence of fission product reactivity effects show that this is dependent upon the data set used and there are significant uncertainties. Some recent thermal reactor studies on approximations in the treatment of decay chains and the importance of xenon and samarium poisoning are also summarized. (author)

  14. Study on the technical feasibility of Fission-Track dating at two irradiation positions of the RA-6 research reactor

    International Nuclear Information System (INIS)

    Dorval, Eric

    2005-01-01

    The method of Fission-Track dating is based upon the detection of the damage caused by fission fragments from the Uranium contained in geological samples.In order to determine the age of a sample, both the amount of spontaneous fissions occurred and the Uranium concentration must be known.The latter requires the irradiation of the samples inside a reactor with a well-thermalized flux, so that fissions are induced over 235 U targets only. Therefore, the Uranium concentration may be determined.The main inconvenient presented by the irradiation sites at the RA-6 MTR-type reactor is that neutron flux is not completely thermal there, which means that fissions due to epithermal and fast neutrons will not be negligible.In the same way, tracks due to fissions of 238 U and 232 Th will be detected. In order to know the corrections that must be applied to those measurements performed in this reactor, it is necessary to characterize fast flux.Because of it, this laboratory's gamma spectrometry equipment had to be calibrated. After that, several activation detectors were irradiated and results were analyzed. Finally, it was determined that it is feasible to Fission-Track date at the I6 position. However, limitations associated to this method were analyzed for the values of flux measured in the different sites

  15. Double stage dry-wet-fermentation - start-up of a pilot biogas plant

    International Nuclear Information System (INIS)

    Buschmann, Jeannette; Busch, Gunter; Burkhardt, Marko

    2009-01-01

    The Brandenburg University of Technology (BTU) has developed a double stage dry-wet fermentation process for fast and safe anaerobic degradation. Originally designed for treatment of organic wastes, this process allows using a wide variety of solid biodegradable materials. The dividing of hydrolysis and methanation in this process, allows an optimization of the different steps of biogas generation separately. The main advantages of the process are the optimum process control, an extremely stable process operation and a high gas productivity and quality. Compared to conventional processes, the retention times within the percolation stage (hydrolysis) are reduced considerably. In cooperation with the engineering and consulting company GICON, the technology was qualified further to an industrial scale. In 2007 a pilot plant, and, simultaneously, an industrial plant were built by GICON based on this double stage technology. Based on practical experience from the operation of laboratory fermentation plants, the commissioning of the pilot plant was planned, controlled and monitored by our institution. The start-up of a biogas plant of this type focuses mainly on the inoculation the of methane reactor. The growth of microbial populations and generation of a stable biocenosis within the methane reactor is essential and affects the duration of starting period as well as the methanation efficiency a long time afterwards. This paper concerns with start-up of a pilot biogas plant and discusses particular occurrences and effects during this period. (author)

  16. Migration of U-series radionuclides around the Bangombe natural fission reactor (Gabon)

    International Nuclear Information System (INIS)

    Bros, R.; Yanase, N.; Isobe, H.; Sato, T.; Iida, Y.; Ohnuki, T.; Roos, P.; Holm, E.

    1999-01-01

    The Bangombe natural fission reactors has undergone extensive weathering phenomena and continues to be affected by the penetration of meteoric waters. Hence this system provides a model for studying the stability of spent fuel uraninite and the influence of various rock matrices on the mobilization/retardation of various actinides and fission products. The Bangombe uranium deposit has been investigated by drilling on a grid. Radiochemical analysis by alpha- and gamma-spectroscopy of the obtained rocks show significant disequilibria of the 234 U/ 238 U, 230 Th/ 234 U, and 226 Ra/ 230 Th parent-daughter pairs. In this paper, a conceptual model for spatio/temporal evolution of the Bangombe system is proposed. (J.P.N.)

  17. The study of two, three and four dimensional nonlinear dynamics of nuclear fission reactors and effective parameters on its behaviour

    International Nuclear Information System (INIS)

    Tajik, M.; Ghasemizad, A.

    2008-01-01

    In this research, new physical fission reactor parameters which have very sensitive effects on the qualitative behavior of a reactor, are introduced. Therefore, the two, the nonlinear dynamics of two, three and four dimensional, considering almost the effective parameters are formulated for describing nuclear fission reactor systems. Using both analytical and numerical methods, the stability and instability of the given dynamical equations and the conditions of stability are studied in these systems. We have shown that the two parameters of the mean energy residence time in fuel and coolant and also their ratios have the most qualitative effects on the dynamical behaviour of a typical nuclear fission reactor. Increasing or decreasing of these parameters from a captain limit can lead to stability or un stability in a given system

  18. Planetary Surface Power and Interstellar Propulsion Using Fission Fragment Magnetic Collimator Reactor

    International Nuclear Information System (INIS)

    Tsvetkov, Pavel V.; Hart, Ron R.; King, Don B.; Rochau, Gary E.

    2006-01-01

    Fission energy can be used directly if the kinetic energy of fission fragments is converted to electricity and/or thrust before turning into heat. The completed US DOE NERI Direct Energy Conversion (DEC) Power Production project indicates that viable DEC systems are possible. The US DOE NERI DEC Proof of Principle project began in October of 2002 with the goal to demonstrate performance principles of DEC systems. One of the emerging DEC concepts is represented by fission fragment magnetic collimator reactors (FFMCR). Safety, simplicity, and high conversion efficiency are the unique advantages offered by these systems. In the FFMCR, the basic energy source is the kinetic energy of fission fragments. Following escape from thin fuel layers, they are captured on magnetic field lines and are directed out of the core and through magnetic collimators to produce electricity and thrust. The exiting flow of energetic fission fragments has a very high specific impulse that allows efficient planetary surface power and interstellar propulsion without carrying any conventional propellant onboard. The objective of this work was to determine technological feasibility of the concept. This objective was accomplished by producing the FFMCR design and by analysis of its performance characteristics. The paper presents the FFMCR concept, describes its development to a technologically feasible level and discusses obtained results. Performed studies offer efficiencies up to 90% and velocities approaching speed of light as potentially achievable. The unmanned 10-tons probe with 1000 MW FFMCR propulsion unit would attain mission velocity of about 2% of the speed of light. If the unit is designed for 4000 MW, then in 10 years the unmanned 10-tons probe would attain mission velocity of about 10% of the speed of light

  19. Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

    2018-01-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

  20. Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

    International Nuclear Information System (INIS)

    Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

    2004-01-01

    Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

  1. Data sheets of fission product release experiments for light water reactor fuel, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo; Yamamoto, Katsumune; Nakazaki, Chozaburo.

    1979-07-01

    This is the second data sheets of fission products (FP) release experiments for light water reactor fuel. Results of five FP release experiments from the third to the seventh are presented: results of pre-examinations of UO 2 pellets, photographs of parts of fuel rod assemblies for irradiation and the assemblies, operational conditions of JMTR and OWL-1, variations of radioiodine-131 level in the main loop coolant during experimental periods, and representative results of post-irradiation examinations of respective fuel rods. (author)

  2. Process for separately recovering uranium, transuranium elements, and fission products of uranium from atomic reactor fuel

    International Nuclear Information System (INIS)

    Balal, A.L.; Metscher, K.; Muehlig, B.; Reichmuth, C.; Schwarz, B.; Zimen, K.E.

    1976-01-01

    Spent reactor fuel elements are dissolved in dilute nitric acid. After addition of acetic acid as a complexing agent, the nitric acid is partly decomposed and the mixture subjected to electrolysis while a carrier liquid, which may be dilute acetic acid or a dilute mixture of acetic acid and nitric acid is caused to flow in the electric field between the electrodes either against the direction of ion migration or transversely thereto. The ions of uranium, plutonium, and other transuranium elements, and of fission products accumulate in discrete portions of the electrolyte and are separately withdrawn as at least three fractions after one or more stages of electrolysis

  3. The effect of the greek research reactor operating schedule on its fission product inventory

    International Nuclear Information System (INIS)

    Annousis, J.N., Armyriotis, J.S.

    1987-12-01

    A simple method to convert the fission product inventory of the 'Democritos' uous Greek Research Reactor (GRR) corresponding to its continuous operation over a given time interval, into the inventory corresponting to GRR discontinuous but periodic operation of the same total duration, is presented in this paper. Relevant correction factors for 31 radioecologically significant radionuclides of the inventory are given as a function of the number of hours or operation per day, 5 days per week of the GRR, according to its present of possible future operating schedule

  4. The effect of the Greek Research Reactor operating schedule on its fission product inventory

    International Nuclear Information System (INIS)

    ANOUSSIS, J.N.; ARMYRIOTIS, J.S.

    1987-12-01

    Full text:A simple method to convert the fission product inventory of ''Demokritos'' Greek Research Reactor(GRR) corresponding to its continuous operation over a given time interval, into the inventory corresponding to GRR discontinuous but periodic operation of the same total duration, is presented in this paper. Relevant correction factors for 31 radioecologically significant radionuclides of the inventory are given as a function of the number of hours of operation per day, 5 days per week of the GRR, according to its present or possible future operating schedule. (author)

  5. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  6. The geo-reactor. A link between nuclear fission and geothermal energy?

    International Nuclear Information System (INIS)

    Degueldre, Claude; Fiorina, Carlo

    2013-01-01

    Recent high-precision isotope analysis data suggests the potential occurrence of a geo-reactor. Specific gas isotopes that could have been generated by binary and ternary fissions were identified in volcano emanations or as soluble/associated species in crystalline rocks and semi-quantitatively evaluated as isotopic ratio or estimated amounts. Presently if it is evident that according to the actinide inventory on the Earth, local potential criticality of the geo-system may have been reached, several questions remain such as why, where and when did a geo-reactor be operational? Even if the hypothesis of a geo-reactor operation in the proto-Earth period should be acceptable, it could be difficult to anticipate that a geo-reactor is still operating today. This could be tested in the future by assessing and reconstructing the system by antineutrino detection and tomography through the Earth. The present paper focuses on the geo-reactor conditions including history, spatial extension and regimes. The discussion based on recent calculations involves investigations on the limits in term of fissile inventory, size and power, based on stratification through the gravitational field and the various features through the inner mantel, the boundary with the core, the external part and the inner-core. the reconstruction allows to formulating that from the history point of view there are possibilities that the geo-reactor reached criticality in a proto-Earth period as a thorium/uranium reactor triggered by an under-layer with heavier actinides. The geo-reactor should be a key component of geothermal energy sources. (author)

  7. Irradiation positions for fission-track dating in the University of Pavia TRIGA Mark II nuclear reactor

    International Nuclear Information System (INIS)

    Oddone, Massimo; Meloni, Sandro; Balestrieri, Maria Laura; Bigazzi, Giulio

    2002-01-01

    An irradiation position arranged is described in the present paper for fission-track dating in the Triga Mark II reactor of the University of Pavia. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively good neutron thermalization (φ th /φ f = 0.956) and lack of significant fluence spatial gradients are good factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author)

  8. Non Nuclear Testing of Reactor Systems In The Early Flight Fission Test Facilities (EFF-TF)

    International Nuclear Information System (INIS)

    Van Dyke, Melissa; Martin, James

    2004-01-01

    The Early Flight Fission-Test Facility (EFF-TF) can assist in the design and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are 'non-nuclear' in nature (e.g. system's nuclear operations are understood). For many systems, thermal simulators can be used to closely mimic fission heat deposition. Axial power profile, radial power profile, and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other Nasa centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004. (authors)

  9. Measurement of fission cross-section of actinides at n_TOF for advanced nuclear reactors

    CERN Document Server

    Calviani, Marco; Montagnoli, G; Mastinu, P

    2009-01-01

    The subject of this thesis is the determination of high accuracy neutron-induced fission cross-sections of various isotopes - all of which radioactive - of interest for emerging nuclear technologies. The measurements had been performed at the CERN neutron time-of-flight facility n TOF. In particular, in this work, fission cross-sections on 233U, the main fissile isotope of the Th/U fuel cycle, and on the minor actinides 241Am, 243Am and 245Cm have been analyzed. Data on these isotopes are requested for the feasibility study of innovative nuclear systems (ADS and Generation IV reactors) currently being considered for energy production and radioactive waste transmutation. The measurements have been performed with a high performance Fast Ionization Chamber (FIC), in conjunction with an innovative data acquisition system based on Flash-ADCs. The first step in the analysis has been the reconstruction of the digitized signals, in order to extract the information required for the discrimination between fission fragm...

  10. A study of fission product transport from failed fuel during N reactor postulated accidents

    International Nuclear Information System (INIS)

    Hagrman, D.L.

    1989-09-01

    This report presents a study of fission product transport behavior in N Reactor during a severe accident. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The current report is an extension to a previous interum study that has added an aerosol formation model, replaced an older aerosol deposition model with an improved correlation, and incorporated results of a revised analysis of the process tubes. The LACE LA1 and LA3 tests are used to assess the revised model applied to determine aerosol deposition. The study concludes that a cesium iodide aerosol is likely to form near the downstream end of the process tubes. Transport of most of the released cesium and iodine as well as less volatile material depends on the behavior of this aerosol and the behavior is sensitive to several parameters that are not well known. If the environment is very clean and effluent flow is sufficient to support oxidation of the zircaloy and uranium of the process tubes, almost none of the aerosol deposits in the riser. Reduction of the effluent flow or the presence of high concentrations of aerosols of very low volatile material like zirconium, uranium, or their oxides causes deposition of the fission products in the riser piping. 24 refs., 18 figs., 11 tabs

  11. New ceramics for nuclear industry. Case of fission and fusion reactors

    International Nuclear Information System (INIS)

    Yvars, M.

    1979-10-01

    The ceramics used in the nuclear field are described as is their behaviour under radiation. 1) Power reactors - nuclear fission. Ceramics enter into the fabrication of nuclear fuels: oxides, carbides, uranium or plutonium nitrides or oxy-nitrides. Silicon carbide SiC is used for preparing the fuels of helium cooled high temperature reactors. Its use is foreseen in the design of gas high temperature gas thermal exchangers, as is silicon nitride (Si 3 N 4 ). In the materials for safety or control rods, the intense neutron flows induce nuclear reactions which increase the temperature of the neutron absorbing material. Boron carbide B 4 C, rare earth oxides Ln 2 O 3 , or B 4 C-Cu or B 4 C-Al cermets are employed. Burnable poison materials are formed of Al 2 O 3 -B 4 C or Al 2 O 3 -Ln 2 O 3 cermets. The moderators of thermal neutron reactors are in high purety polycrystalline graphite. For the thermal insulation of reactor vessels and jackets, honeycomb ceramics are used as well as ceramic fibres on an increasing scale (kaolin, alumina and other fibres). 2) fusion reactors (Tokomak). These require refractory materials with a low atomic number. Carbon fibres, boron carbide, some borons (Al B 12 ), silicon nitrides and oxy-nitrides and high density alumina are the substances considered [fr

  12. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  13. Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors (Workshop Report)

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, RE

    2004-07-15

    The ''Workshop on Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors'' was convened to determine the degree to which an increased effort in modeling and simulation could help bridge the gap between the data that is needed to support the implementation of these advanced nuclear technologies and the data that can be obtained in available experimental facilities. The need to develop materials capable of performing in the severe operating environments expected in fusion and fission (Generation IV) reactors represents a significant challenge in materials science. There is a range of potential Gen-IV fission reactor design concepts and each concept has its own unique demands. Improved economic performance is a major goal of the Gen-IV designs. As a result, most designs call for significantly higher operating temperatures than the current generation of LWRs to obtain higher thermal efficiency. In many cases, the desired operating temperatures rule out the use of the structural alloys employed today. The very high operating temperature (up to 1000 C) associated with the NGNP is a prime example of an attractive new system that will require the development of new structural materials. Fusion power plants represent an even greater challenge to structural materials development and application. The operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable to or greater than those for proposed Gen-IV fission reactors. In addition, the transmutation products created in the structural materials by the high energy neutrons produced in the DT plasma can profoundly influence the microstructural evolution and mechanical behavior of these materials. Although the workshop addressed issues relevant to both Gen-IV and fusion reactor materials, much of the discussion focused on fusion; the same focus is reflected in this report. Most of the physical models and computational methods

  14. The different facilities of the reactor Phenix for radio isotope production and fission product burner

    International Nuclear Information System (INIS)

    Coulon, P.; Clerc, R.; Tommasi, J.

    1993-01-01

    During the last few years different tests have been made to optimize the blanket of the reactor. Year after year the breeding ratio has lost a part of interest regarding the production and availability of plutonium in the world. A characteristic of a fast reactor is to have important neutron leaks from the core. The spectrum of those neutrons is intermediate, the idea was to find a moderator compatible with sodium and stable in temperature. After different tests we kept as a moderator the calcium hydride and as a samply support, a cluster which is separated from the carrier. At the end we present the model used for thermalized calculations. The scheme is then applied to a heavy nuclide transmutation example (Np237 Pu238) and to fission product transmutation (Tc99). (authors). 9 figs

  15. 99Tc, Pb and Ru migration around the Oklo natural fission reactors

    International Nuclear Information System (INIS)

    Gancarz, A.; Cowan, G.; Curtis, D.; Maeck, W.

    1980-01-01

    This work demonstrates the utility of the Oklo uranium ore deposit and natural fission reactors as a long time scale analogue for man-made radioactive waste repositories. It has been shown that the ores and nearby rocks were open to the loss and gain of 99 Tc, ruthenium, and lead relative to uranium. Identified regions of element deficiencies and those which are correspondingly enriched are separated by less than 10 meters. However, more extensive sampling is required to define the overall extent of the element migration. Element fractionation took place on at least two vastly different time scales; 99 Tc was fractionated from ruthenium within one million years of the end of reactor criticality. Lead-uranium fractionation has been ongoing for most of the two billion years since the ores were formed. Diffusion loss of lead from host uraninite appears to be an important process in the fractionation of lead from uranium

  16. Transmutation of long-lived fission product (137Cs, 90Sr) by a reactor-accelerator system

    International Nuclear Information System (INIS)

    Toyama, Shin-ichi; Takashita, Hirofumi; Konashi, Kenji; Sasao, Nobuyuki; Sato, Isamu.

    1990-01-01

    The report discusses the transmutation of long-lived fission products by a reactor and accelerator. It is important to take some criteria into consideration in transmutation disposal. To satisfy the criteria, a combined system of a reactor and an accelerator is proposed for the transmutation. An outline of the transmutation reactor and the accelerator is presented. The transmutation reactor has the ability to transmute a large quantity of fission products. However, it is desirable to have a high transmutation rate as well as a large disposal ability. Besides the transmutation property, it is necessary to investigate the physics of the transmutation reactor such as nuclear characteristics and burnup properties in order to obtain the most suitable, high performance core concept. A study on those properties is also presented. A high power accelerator is required for the transmutation. So a test linac is developed to accelerate high intensity beams. (N.K.)

  17. Neutronics analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Jiang Jieqiong; Wang Minghuang; Chen Zhong; Qiu Yuefeng; Liu Jinchao; Bai Yunqing; Chen Hongli; Hu Yanglin

    2010-01-01

    Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1 GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33 000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication.

  18. Lithium Hideout and Return in the CANDU Heat Transport System during Shutdown and Start-up

    International Nuclear Information System (INIS)

    Qiu, L.; Snaglewski, A.P.

    2012-09-01

    Lithium hydroxide is used to control the pH a (pH apparent) of the Heat Transport System (HTS) coolant in CANDU R reactors. The recommended range of the lithium concentration in the coolant is between 0.38 ppm (5.5x10 -5 m) and 0.60 ppm (8.7x10 -5 m) to minimize carbon steel corrosion in the HTS and magnetite deposition in the core during normal operation; this corresponds to pH a values between 10.2 and 10.4. Similar pH a and lithium concentrations should be maintained during shutdown and start-up. However, maintaining the pH a of the HTS coolant within specification during shutdown and start-up has been difficult for some CANDU stations, especially when the HTS is taken to a Low Level Drain State (LLDS), because of lithium hideout and return. This paper presents the results from lithium adsorption and desorption studies on iron oxides under relevant shutdown and start-up chemistry conditions performed to elucidate the mechanisms of the observed lithium hideout and return. The results show that lithium hideout and return are driven largely by changes in the solubility of magnetite as the HTS coolant chemistry changes during shutdown; changes in lithium concentration were inversely correlated with the solubility of magnetite. When the HTS system is de-pressurized and drained to a low coolant level, the ingress of air rapidly oxidizes the dissolved Fe (II) in the coolant, 2Fe +2 + 1 / 2 O 2 + 3 H 2 = 2FEOOH + 4 H + , resulting in the formation of lepidocrocite or maghemite, which have much lower solubilities but larger surface areas than does magnetite. The large surface area of the Fe (III) oxides can adsorb significant quantities of lithium from the coolant, leading to lithium hideout and a pH a decrease. During start-up, the chemistry of the coolant changes from oxidizing to reducing, and lepidocrocite and other Fe (III) oxides are reduced to Fe (II), gradually dissolving as their solubility increases with increasing temperature. The adsorbed lithium is released

  19. Evaluation of transmutation performance of long-lived fission products with a super fast reactor

    International Nuclear Information System (INIS)

    Lu, Haoliang; Han, Chiyoung; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    The performance of the Super Fast Reactor for transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super Fast Reactor. First is in the blanket assembly due to the ZrH 1.7 layer which can slow down the fast neutrons. Second is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected of transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe·y and 2.79%/GWe.y can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the outputs from 11.8 and 6.2 1000MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained by the Super Fast Reactor. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super Fast Reactor. It turns out that the 135 Cs transmutation is a challenge not only for the Super Fast Reactor but also for other commercial fast reactors. (author)

  20. Start-up and operation of Laguna Verde-2

    International Nuclear Information System (INIS)

    Torres Ramirez, J.F.

    1996-01-01

    The 665 MWe Laguna Verde-2 nuclear generating unit was accepted into commercial operation on April 10, 1995. The boiling water reactor plant by General Electric (GE) was first synchronized with the grid on November 11, 1994. Laguna Verde-2 is identical with Laguna Verde-1 on the same site. That unit had gone critical for the first time in November 1988 and had first been synchronized with the power grid on April 13, 1989. Commercial operation of Laguna Verde-1 had been started on July 29, 1990. Mexico's only nuclear power plant had been built 70 km north of Veracruz on the east coast and had been scheduled to start operation in 1976. As the Mexican nuclear power program was reduced, the scheduled commissioning dates suffered more and more delays. In the full of 1987, the investigation by the Operational Safety Review Team (Osart) of the International Atomic Energy Agency (IAEA) had indicated that the safety requirements of installations and equpiment were met, and that the whole plant was well prepared for fueling. In mid-1988, the Mexican Government had issued the permit to fuel the Laguna Verde-1 reactor. The contract to build the two units had been awarded in 1972/73. No other nuclear power plants are currently under construction or in the planning phase in Mexico. (orig.) [de

  1. 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)

  2. Irradiation effects in fused quartz 'Suprasil' as a detector of fission fragments under high flux of reactor neutrons

    International Nuclear Information System (INIS)

    Moraes, O.M.G. de.

    1984-01-01

    A systematic study about the registration characteristics of synthetic fused quartz 'Suprasil I' use as a detector of fission fragments under high flux of reactor neutrons and the effects of irradiation on it was performed. Fission fragments of 252 Cf, gamma radiation doses of of 60 Co up to 150 MGy, and integrated neutrons fluxes up to 10 20 n/cm 2 were used. A model to explain the effects on track registration and development characteristics of 'Suprasil I' irradiated on reactors were proposed, based on the obtained results for efficiency an for annealing. (C.G.C.) [pt

  3. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    International Nuclear Information System (INIS)

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-01-01

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  4. The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

    International Nuclear Information System (INIS)

    Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

    2012-01-01

    In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

  5. Modelling and simulation the radioactive source-term of fission products in PWR type reactors

    International Nuclear Information System (INIS)

    Porfirio, Rogilson Nazare da Silva

    1996-01-01

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  6. Aloittelevan start-up yrittäjän opas rahoitukseen

    OpenAIRE

    Lamminsalo, Matti

    2015-01-01

    Opinnäytetyö käsittelee start-up rahoitusta ja siihen liittyviä toimintamalleja. Opinnäytetyön tuloksena on syntynyt opaskirja: Aloittelevan start-up yrittäjän opas rahoitukseen. Research about start-up ecosystem with investment aspect. As a result guidebook for start-up investment and how to raise a fund was writen.

  7. Dynamics and Control During Start-Up of Energy Integrated Distillation Column

    DEFF Research Database (Denmark)

    of the inherent effects of process integration upon a start-up procedure. In particular the effect of energy recycle upon the possible start-up sequence, and the effect of using an actuator for control purposes, i.e. the heat pump, which in itself requires significant start-up time. The experimental application...

  8. Dynamics and Control During Start-up of Energy Integrated Distillation Column

    DEFF Research Database (Denmark)

    Eden, Mario Richard; Koggersbøl, Arne; Hallager, L.

    2000-01-01

    of the inherent effects of process integration upon a start-up procedure. In particular the effect of energy recycle upon the possible start-up sequence, and the effect of using an actuator f or control purposes, i.e. the heat pump, which in itself requires significant start-up time. The experimental application...

  9. 26 CFR 1.195-1T - Election to amortize start-up expenditures (temporary).

    Science.gov (United States)

    2010-04-01

    ... business, or $5,000 (reduced (but not below zero) by the amount by which the start-up expenditures exceed... beginning with the month in which the active trade or business begins. All start-up expenditures that relate to the active trade or business are considered in determining whether the start-up expenditures...

  10. 78 FR 73753 - Partnerships; Start-Up Expenditures; Organization and Syndication Fees

    Science.gov (United States)

    2013-12-09

    ... deduct start-up expenditures in the taxable year in which the active trade or business begins. The amount... begins. All start-up expenditures that relate to the active trade or business are considered in...(b) to amortize start-up expenditures for the taxable year in which the active trade or business to...

  11. Natural fission reactors from Gabon. Contribution to the study of the conditions of stability of a natural radioactive wastes storage site (2 Ga)

    International Nuclear Information System (INIS)

    Pourcelot, L.

    1997-01-01

    The natural fission reactors of Oklo consists of a core of uraninite (60%) with fission products, embedded in a pure clay matrix. Thus, the aim of geological, mineral, and geochemical studies of the Oklo Reactors is to assess the behaviour of fission products in an artificial waste depository. Previous studies have shown that Reactor Zone 10, located in the Oklo mine, represents an example for an exceptional confinement of fission products since 2 Ga. In reactor Zone 9, located in Oklo open pit, migrations are more important. Reactor ZOne 13 was influenced by a thermal event due to a doleritic intrusion, located some twenty meters far away, one Ga years after fission reaction operations. In this study,we characterized temperature and redox conditions of fluids by using stable isotopes of uraninites and clays. Moreover mineralogical and chemical characteristics were defined. (author)

  12. Method for heating of the primary circuit of WWER electric power units at cold start-up

    International Nuclear Information System (INIS)

    Ivanov, I.N.; Dimitrov, B.D.; Korkinova, M.I.

    1982-01-01

    The method increases the heating rate and shorten the start-up time of the electric power units. It comprises a primary stopping of the reactor core heating and provides a forced circulation of the heat-carrier through the circulation cycles of the primary circuit. The thermal energy is supplied in one or several steam generators in the secondary circuit of an NPP operating unit. 1 cl., 3 figs

  13. Results of physics start-up tests of Mochovce and Bohunice units with 2-nd generation Gd fuel (average enrichment 4.87 %)

    International Nuclear Information System (INIS)

    Polakovic, F.

    2015-01-01

    There are presented main features of the fuel and the list of experimental neutron-physical characteristics measured during physics start-up tests.All together there were carried out 14 physics start-ups at Bohunice and Mochovce Units with the new type of fuel. Differences between theoretical and experimental neutron-physical characteristics were statistically processed and compared with the tests acceptance criteria. There are summarized results of reactor physics start-ups with 2-nd generation Gd fuel usage [ru

  14. NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stevenson, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tsai, Kevin [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    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 fission chambers 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 two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating.

  15. Development of Start-up and Shutdown Procedure for the HANARO Fuel Test Loop

    International Nuclear Information System (INIS)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, J. M.; Lee, C. Y.; Ahn, S. H.

    2009-06-01

    A start-up and shutdown procedure for the HANARO fuel test loop has been developed. This is a facility for fuel and material irradiation tests. The facility provides experimental conditions similar to the normal operational pressures and temperatures of commercial PWR and CANDU plants. The normal operation modes of the HANARO fuel test loop are classified into loop shutdown, cold stand-by 1, cold stand-by 2, hot stand-by, and hot operation. The operation modes depend on the fission power of test fuels and the coolant temperature at the inlet of the in-pile test section. The HANARO must maintain a shutdown mode if the HANARO fuel test loop is loop shutdown, cold stand-by 1, cold stand-by 2, or hot stand-by. As the HANARO becomes power operation mode, the operation mode of the HANARO fuel test loop comes to hot operation from hot stand-by. The procedure for the HANARO fuel test loop consists of four main parts such as check of initial conditions, stat-up operation procedure, shutdown operation procedure, and check lists for operations. Several hot test operations ensure that the procedure is appropriate

  16. 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

  17. Start-Up Size Strategy and Risk Management: Impact on New Venture Performance

    OpenAIRE

    André van Stel; Andrew Burke; José Maria Millán; Concepcion Roman

    2013-01-01

    Start-up size is a key strategic decision for entrepreneurs. Should entrepreneurs start up close to minimum efficient scale or should they take less risks and start-up on a smaller scale? Previously, this strategic decision appeared to be one of simply making a choice between a higher risk/reward larger start-up versus a lower risk/reward smaller scale start-up. However, recent research on the relationship between risk management and performance (Burke, 2009) indicates that in situations of g...

  18. Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

    International Nuclear Information System (INIS)

    Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

    2009-01-01

    Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

  19. Computation of fission product distribution in core and primary circuit of a high temperature reactor during normal operation

    International Nuclear Information System (INIS)

    Mattke, U.H.

    1991-08-01

    The fission product release during normal operation from the core of a high temperature reactor is well known to be very low. A HTR-Modul-reactor with a reduced power of 170 MW th is examined under the aspect whether the contamination with Cs-137 as most important nuclide will be so low that a helium turbine in the primary circuit is possible. The program SPTRAN is the tool for the computations and siumlations of fission product transport in HTRs. The program initially developed for computations of accident events has been enlarged for computing the fission product transport under the conditions of normal operation. The theoretical basis, the used programs and data basis are presented followed by the results of the computations. These results are explained and discussed; moreover the consequences and future possibilities of development are shown. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  1. Start-up phase of a two-stage anaerobic co-digestion process: Hydrogen and methane production from food waste and vinasse from ethanol industry

    DEFF Research Database (Denmark)

    Náthia-Neves, Grazielle; Neves, Thiago de Alencar; Berni, Mauro

    2018-01-01

    The start-up conditions of mesophilic anaerobic co-digestion of restaurant food waste and vinasse, a waste from sugarcane industry, was investigated for efficient biogas production. A pilot plant, containing two reactors, was designed and used sequentially and semi-continuously for biogas...... production. All effective operational parameters were controlled in both reactors over the course of the study. The results indicated that the organic matters were quickly decreased during the start-up phase in the first reactor, resulting in 52% and 64% reduction in total solids and total volatile solids...

  2. Development and optimization of neutron measurement methods by fission chamber on experimental reactors - management, treatment and reduction of uncertainties

    International Nuclear Information System (INIS)

    Blanc-De-Lanaute, N.

    2012-01-01

    The main objectives of this research thesis are the management and reduction of uncertainties associated with measurements performed by means of a fission-chamber type sensor. The author first recalls the role of experimental reactors in nuclear research, presents the various sensors used in nuclear detection (photographic film, scintillation sensor, gas ionization sensor, semiconducting sensor, other types of radiation sensors), and more particularly addresses neutron detection (activation sensor, gas filling sensor). In a second part, the author gives an overview of the state of the art of neutron measurement by fission chamber in a mock-up reactor (signal formation, processing and post-processing, associated measurements and uncertainties, return on experience of measurements by fission chamber on Masurca and Minerve research reactors). In a third part, he reports the optimization of two intrinsic parameters of this sensor: the thickness of fissile material deposit, and the pressure and nature of the filler gas. The fourth part addresses the improvement of measurement electronics and of post-processing methods which are used for result analysis. The fifth part deals with the optimization of spectrum index measurements by means of a fission chamber. The impact of each parameter is quantified. Results explain some inconsistencies noticed in measurements performed on the Minerve reactor in 2004, and allow the improvement of biases with computed values [fr

  3. Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

    Science.gov (United States)

    Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

    2017-10-24

    Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

  4. Lanthanide fission product separation from the transuranics in the integral fast reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Ackerman, J.P.

    1993-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed by Argonne National Laboratory. This reactor uses liquid-metal cooling and metallic fuel. Its spent fuel will be reprocessed using a pyrochemical method employing molten salts and liquid metals in an electrofining operation. The lanthanide fission products are a concern during reprocessing because of heating and fuel performance issues, so they must be removed periodically from the system to lessen their impact. The actinides must first be removed form the system before the lanthanides are removed as a waste stream. This operation requires a relatively good lanthanide-actinide separation to minimize both the amount of transuranic material lost in the waste stream and the amount of lanthanides collected when the actinides are first removed. A computer code, PYRO, that models these operations using thermodynamic and empirical data was developed at Argonne and has been used to model the removal of the lanthanides from the electrorefiner after a normal operating campaign. Data from this model are presented. The results demonstrate that greater that 75% of the lanthanides can be separated from the actinides at the end of the first fuel reprocessing campaign using only the electrorefiner vessel

  5. Fusion technology development: role of fusion facility upgrades and fission test reactors

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.

    1983-01-01

    The near term national fusion program is unlikely to follow the aggressive logic of the Fusion Engineering Act of 1980. Faced with level budgets, a large, new fusion facility with an engineering thrust is unlikely in the near future. Within the fusion community the idea of upgrading the existing machines (TFTR, MFTF-B) is being considered to partially mitigate the lack of a design data base to ready the nation to launch an aggressive, mission-oriented fusion program with the goal of power production. This paper examines the cost/benefit issues of using fusion upgrades to develop the technology data base which will be required to support the design and construction of the next generation of fusion machines. The extent of usefulness of the nation's fission test reactors will be examined vis-a-vis the mission of the fusion upgrades. The authors show that while fission neutrons will provide a useful test environment in terms of bulk heating and tritium breeding on a submodule scale, they can play only a supporting role in designing the integrated whole modules and systems to be used in a nuclear fusion machine

  6. Fusion technology development: role of fusion facility upgrades and fission test reactors

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Miller, L.G.; Longhurst, G.R.; Schmunk, R.E.

    1983-01-01

    The near term national fusion program is unlikely to follow the aggressive logic of the Fusion Engineering Act of 1980. Faced with level budgets, a large, new fusion facility with an engineering thrust is unlikely in the near future. Within the fusion community the idea of upgrading the existing machines (TFTR, MFTF-B) is being considered to partially mitigate the lack of a design data base to ready the nation to launch an aggressive, mission-oriented fusion program with the goal of power production. This paper examines the cost/benefit issues of using fusion upgrades to develop the technology data base which will be required to support the design and construction of the next generation of fusion machines. The extent of usefulness of the nation's fission test reactors will be examined vis-a-vis the mission of the fusion upgrades. We will show that while fission neutrons will provide a useful test environment in terms of bulk heating and tritium breeding on a submodule scale, they can play only a supporting role in designing the integrated whole modules and systems to be used in a nuclear fusion machine

  7. Local Fission Gas Release and Swelling in Water Reactor Fuel during Slow Power Transients

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg; Walker, C.T.; Ray, I.L.F.

    1985-01-01

    Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7–4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25–40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two...... factors that determine the level of fission gas release during a power bump. Release begins when gas bubbles on grain boundaries start o interlink. This occurred at r/r0 ~ 0.75. Release tunnels were fully developed at r/r0 ~ 0.55 with the result that gas release was 60–70% at this position....

  8. Continuous radiochemical analysis of fission products in a nuclear reactor water coolant

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Zakharov, L.K.; Leont'ev, G.G.; Mel'nikov, V.A.; Orlenkov, I.S.; Slutskij, G.K.

    1975-01-01

    Method for continuous radiochemical analysis of I, Cs, Ba, Sr and Ce isotopes in a reactor water heat-transfer agent was developed. A continuous two-dimensional chromatographic process of complex mixtures separation of substances proved to be feasible on several parallel sorbent layers, which moved at constant velocities and separated by stationary intermediate collectors. Tests on model solutions containing I, Ce, Cs and Ba isotopes and on heat-carrier samples showed quantitative separation of elements. The results were indicative of a basic possibility of using multisorbent chromatographs for continuous control of multicomponent mixtures, particularly for control of radioactive fission product compositions in water heat-transfer agents in nuclear power plants. A diagram is shown for a two-dimensional chromatographic separation of a multicomponent mixture. Also shown is a flow chart of an installation for continuous control of iodine and cesium isotope activities

  9. Overview of Fusion-Fission Hybrid Reactor Design Study in China

    International Nuclear Information System (INIS)

    Huang Jinhua; Feng Kaiming; Deng Baiquan; Deng, P.Zh.; Zhang Guoshu; Hu Gang; He Kaihui; Wu Yican; Qiu Lijian; Huang Qunying; Xiao Bingjia; Liu Xiaoping; Chen Yixue; Kong, M.H.

    2002-01-01

    The motivation for developing fusion-fission hybrid reactors is discussed in the context of electricity power requirements by 2050 in China. A detailed conceptual design of the Fusion Experimental Breeder (FEB) was developed from 1986-1995. The FEB has a subignited tokamak fusion core with a major radius of 4.0 m, a fusion power of 145 MW, and a fusion energy gain Q of 3. Based on this, an engineering outline design study of the FEB, FEB-E, has been performed. This design study is a transition from conceptual to engineering design in this research. The main results beyond that given in the detailed conceptual design are included in this paper, namely, the design studies of the blanket, divertor, test blanket, and tritium and environment issues. In-depth analyses have been performed to support the design. Studies of related advanced concepts such as the waste transmutation blanket concept and the spherical tokamak core concept are also presented

  10. Dosimetry of fission neutrons in a 1-W reactor, UTR-KINKI

    CERN Document Server

    Endo, S; Yoshitake, Y

    2002-01-01

    The energy spectrum of fission neutrons in the biological irradiation field of the Kinki University reactor, UTR-KINKI, has been determined by a multi-foil activation analysis coupled with artificial neural network techniques and a Au-foil activation method. The mean neutron energy was estimated to be 1.26+-0.05 MeV from the experimentally determined spectrum. Based on this energy value and other information, the neutron dose rate was estimated to be 19.7+-1.4 cGy/hr. Since this dose rate agrees with that measured by a pair of ionizing chambers (21.4 cGy/hr), we conclude that the mean neutron energy could be estimated with reasonable accuracy in the irradiation field of UTR-KINKI. (author)

  11. Cavity Ring-Down Spectroscopy for Gaseous Fission Products Trace Measurements in Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Jacquet, P.; Pailloux, A.; Doizi, D.; Aoust, G.; Jeannot, J.-P.

    2013-06-01

    Safety and availability are key issues of the generation IV reactors. Hence, the three radionuclide confinement barriers, including fuel cladding, must stay tight during the reactor operation. During the primary gaseous failure, fission products xenon and krypton are released. Their fast and sensitive detection guarantees the first confinement barrier tightness. In the frame of the French ASTRID project, an optical spectroscopy technique - Cavity Ring Down Spectroscopy (CRDS) - is investigated for the gaseous fission products measurement. A dedicated CRDS set-up is needed to detect the rare gases with a commercial laser. Indeed, the CRDS is coupled to a glow discharge plasma, which generates a population of metastable atoms. The xenon plasma conditions are optimized to 110 Pa and 1.3 W (3 mA). The production efficiency of metastable Xe is then 0.8 %, stable within 0.5% during hours. The metastable number density is proportional to the xenon over argon molar fraction. The spectroscopic parameters of the strong 823.16 nm xenon transition are calculated and/or measured in order to optimize the fit of the experimental spectra and make a quantitative measurement of the metastable xenon. The CRDS is coupled to the discharge cell. The laser intensity inside the cavity is limited by the optical saturation process, resulting from the strong optical pumping of the metastable state. The resulting weak CRDS signal requires a fast and very sensitive photodetector. A 600 ppt xenon molar fraction was measured by CRDS. With the present set-up, the detection limits are estimated from the baseline noise to approximately 20 ppt for each even isotope, 60 ppt for the 131 Xe and 55 ppt for the 129 Xe. This sensitivity matches the specifications required for gaseous leak measurement; approximately 100 ppt for 133 Xe (4 GBq/m 3 ) and 10 ppb for stable isotopes. The odd isotopes are selectively measured, whereas the even isotopes overlap, a spectroscopic feature that applies for stable or

  12. Established Companies meet Start-ups : Case: A service to research developing business models in start-up hot spots around the globe

    OpenAIRE

    Eisenblätter, Simon

    2016-01-01

    This thesis presents a business model with the aim to improve conversation between traditional companies and newly established start-ups. The assumption that such conversation can be facilitated by scouting for innovation in start-up hot spots around the world, will be discussed throughout the paper. The purpose of this work is to lay out a plan which can be presented to a first range of possible customers, with the goal of negotiating a trial cooperation. Before introducing the busi...

  13. Power deposition distribution in liquid lead cooled fission reactors and effects on the reactor thermal behaviour

    International Nuclear Information System (INIS)

    Cevolani, S.; Nava, E.; Burn, K. W.

    2001-01-01

    In the framework of an ADS study (Accelerator Driven System, a reactor cooled by a lead bismuth alloy) the distribution of the deposited energy between the fuel, coolant and structural materials was evaluated by means of Monte Carlo calculations. The energy deposition in the coolant turned out to be about four percent of the total deposited energy. In order to study this effect, further calculations were performed on water and sodium cooled reactors. Such an analysis showed, for both coolant materials, a much lower heat deposition, about one percent. Based on such results, a thermohydraulic analysis was performed in order to verify the effect of this phenomenon on the fuel assembly temperature distribution. The main effect of a significant fraction of energy deposition in the coolant is concerned with the decrease of the fuel pellet temperature. As a consequence, taking into account this effect allows to increase the possibilities of optimization at the disposal of the designer [it

  14. Comparison of 252Cf time correlated induced fission with AmLi induced fission on fresh MTR reserach reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jay Prakash [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-01

    The objectives of this project are to calibrate the Advanced Experimental Fuel Counter (AEFC), benchmark MCNP simulations using experimental results, investigate the effects of change in fuel assembly geometry, and finally to show the boost in doubles count rates with 252Cf active soruces due to the time correlated induced fission (TCIF) effect.

  15. Construction, start-up and operation of Dukovany nuclear power plant

    International Nuclear Information System (INIS)

    Hornicek, Z.

    1989-01-01

    The Labor Safety Inspectorate have been supervising the construction of the Dukovany nuclear power plant since the construction start in 1977. It was found that in concreting the reactor building walls with concrete mixes, the regulations were not observed on the highest level for the concrete mix drop, and on the gap processing and concrete treatment when concreting was interrupted. Thus, concreting was halted until the conditions for concreting were met. Attention was focused on the protection of the hermetic casing from damage, which had very often happened. A number of shortcomings were detected in storing technology parts. The cleanliness was inspected of the facilities being assembled. The inspections also revealed shortcomings in sealed space tightness. The inspections of assembly and testing of facilities showed failures of the facilities themselves (control valves, electric motors, filter and pump sealing) and of the assembly process. Faults were also detected in electrical equipment. Only a very small part of the installation showed lifetime as specified by Decree 105/1982 Coll. laws on safety assurance in the nuclear power industry. Missing data in documentation led to delays in the start-up stages. The State Surveillance Body also inspected the results of equipment testing. Prior to physical start-up, all production facilities and buildings were inspected, labor safety was inspected for all personnel in communicating corridors, staircases, manholes and onservice and handling posts. Shortcomings were removed. The organization of assembly work was a considerable problem if staff from more organizations had to work together in the same workplace. A list of tasks is presented of State Surveillance Body in operation, maintenance, outages, repairs, and troubleshooting in a nuclear power plant. (J.B.)

  16. Radiochemical studies on fission

    Energy Technology Data Exchange (ETDEWEB)

    None

    1973-07-01

    Research progress is reported on nuclear chemistry; topics considered include: recoil range and kinetic energy distribution in the thermal neutron ftssion of /sup 245/Cm; mass distribution and recoil range measurements in the reactor neutron-induced fission of /sup 232/U; fission yields in the thermal neutron fission of /sup 241/PU highly asymmetric binary fission of uranium induced by reactor neutrons; and nuclear charge distribution in low energy fission. ( DHM)

  17. Preparations for the start-up of a research program in nuclear safeguards at Chalmers

    International Nuclear Information System (INIS)

    Avdic, Senada; Pazsit, Imre

    2004-03-01

    The Department of Reactor Physics at Chalmers University of Technology plans to start-up a research program in nuclear safeguards and nuclear material management. The program is aimed at utilizing the experimental facilities as well as the experience in reactor physics, criticality safety, signal processing and unfolding, and experimental nuclear techniques, in tackling problems in non-destructive assay (NDA) of nuclear materials. For the introductory part of this program, support has been received from the Swedish Nuclear Power Inspectorate to host Dr. Senada Avdic, University of Tuzla, Bosnia, as a post-doc for three months to participate in the preparatory program. The preparations were focussed on a survey of existing active non-destructive assay methods and preparations of their application in the experimental and theoretical/calculational research of our Department. The methods surveyed comprise - the use of a 252 Cf source in active NDA measurements; - planning of an experiment with the existing equipments of the Department; - time correlation measurements with a 252 Cf source and/or a 252 Cf detector; - Monte Carlo simulations of the time correlations between gammas and neutrons from a measurement with a 252 Cf detector: the MCNP-PoliMi code; - Identification of fissile material (enrichment/mass) with 252 Cf measurements; the use of various unfolding techniques (artificial neural networks) for identifying nuclear parameters; use of neutron activation analysis with a neutron generator for determination of distribution of material in an unknown sample; - determination of fissile material content by measurements of delayed neutrons

  18. Start-Up and Aeration Strategies for a Completely Autotrophic Nitrogen Removal Process in an SBR

    Directory of Open Access Journals (Sweden)

    Xiaoling Zhang

    2017-01-01

    Full Text Available The start-up and performance of the completely autotrophic nitrogen removal via nitrite (CANON process were examined in a sequencing batch reactor (SBR with intermittent aeration. Initially, partial nitrification was established, and then the DO concentration was lowered further, surplus water in the SBR with high nitrite was replaced with tap water, and continuous aeration mode was turned into intermittent aeration mode, while the removal of total nitrogen was still weak. However, the total nitrogen (TN removal efficiency and nitrogen removal loading reached 83.07% and 0.422 kgN/(m3·d, respectively, 14 days after inoculating 0.15 g of CANON biofilm biomass into the SBR. The aggregates formed in SBR were the mixture of activated sludge and granular sludge; the volume ratio of floc and granular sludge was 7 : 3. DNA analysis showed that Planctomycetes-like anammox bacteria and Nitrosomonas-like aerobic ammonium oxidization bacteria were dominant bacteria in the reactor. The influence of aeration strategies on CANON process was investigated using batch tests. The result showed that the strategy of alternating aeration (1 h and nonaeration (1 h was optimum, which can obtain almost the same TN removal efficiency as continuous aeration while reducing the energy consumption, inhibiting the activity of NOB, and enhancing the activity of AAOB.

  19. Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

    International Nuclear Information System (INIS)

    Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

    2000-01-01

    Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

  20. Start-up Success in a Small Island State: A Study among Entrepreneurs in Malta

    Directory of Open Access Journals (Sweden)

    Leonie Baldacchino

    2008-05-01

    Full Text Available This study focuses on entrepreneurs in the small island state of Malta and investigates whether starting up and running an enterprise is facilitated or hindered by being in a small island environment. Specifically it asks (1 whether being on a small island, on the periphery of a major market facilitates or hinders entrepreneurship and start-up success; (2 whether Malta’s cultural context and enterprise environment affect entrepreneurship and start-up success; (3 what the key success factors among Maltese start-ups are; and (4 how are creativity and innovation reflected in Maltese start-ups. Qualitative research among 13 start-ups is supported by telephone-based research among a sample of 90 respondents. Findings contribute to the pool of business expertise and context-specific information from small island states that is often missing from the international literature.