WorldWideScience

Sample records for source ans reactor

  1. 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

    reactors. The preliminary results suggest that initial operation in D–D with continual feedback into the plasma of the tritium produced enables a fusion reactor designed solely for D–T operation to start-up in an acceptably short time-scale without the need for any external tritium source.

  2. Advanced energy system with nuclear reactors as an energy source

    International Nuclear Information System (INIS)

    Kato, Y.; Ishizuka, T.; Nikitin, K.

    2007-01-01

    About two-thirds of the energy generated in a light water reactors (LWRs) core is currently dissipated to the ocean as lukewarm water through steam condensers; more than half the energy in helium (He) gas turbine high temperature gas cooled reactors (HTGRs) is dissipated through pre-coolers and inter coolers. The new waste heat recovery system efficiently recovers the waste heat from reactors using boiling heat transfer of 20 degree C liquid carbon dioxide (CO 2 ) instead of conventional sea water as a cooling medium. The CO 2 gasified in the cooling process is used directly as a working fluid of mechanical heat pumps for hot water supply. In LWRs, the net energy utilization fraction to total heat generation in the core exceeds 85% through the waste heat recovery. This cogeneration system is about 2.5 times more effective than current systems in reducing global warming gas emissions and long half- life radioactive material accumulation. It also increases uranium resource utilization relative to current LWRs. In the HTGR cogeneration system, the waste heat is also useful for cold water supply by introducing an adsorption refrigeration system since the gas temperature is still as high as about 190 degree Celsius. When the heat recovery system is incorporated into the HTGR, the electricity to heat-supply ratio of the HTGR cogeneration system accommodates the demand ratio in cities well; it would be suited to dispersed energy sources. The heat supply cost is expected to be lower than those of conventional fossil-fired boilers beyond operation of about four years. The waste heat recovered is able to be utilized not only for local heat supply but also for methane and methanol production from waste products of cities and farms through high-temperature fermentation, e.g., garbage, waste wood and used paper that are produced in cities, along with excreta produced through farming. The methane and methanol can be used to generate hydrogen for fuel cells. The new waste heat

  3. An Open Source-based Approach to the Development of Research Reactor Simulator

    International Nuclear Information System (INIS)

    Joo, Sung Moon; Suh, Yong Suk; Park, Cheol Park

    2016-01-01

    In reactor design, operator training, safety analysis, or research using a reactor, it is essential to simulate time dependent reactor behaviors such as neutron population, fluid flow, and heat transfer. Furthermore, in order to use the simulator to train and educate operators, a mockup of the reactor user interface is required. There are commercial software tools available for reactor simulator development. However, it is costly to use those commercial software tools. Especially for research reactors, it is difficult to justify the high cost as regulations on research reactor simulators are not as strict as those for commercial Nuclear Power Plants(NPPs). An open source-based simulator for a research reactor is configured as a distributed control system based on EPICS framework. To demonstrate the use of the simulation framework proposed in this work, we consider a toy example. This example approximates a 1-second impulse reactivity insertion in a reactor, which represents the instantaneous removal and reinsertion of a control rod. The change in reactivity results in a slightly delayed change in power and corresponding increases in temperatures throughout the system. We proposed an approach for developing research reactor simulator using open source software tools, and showed preliminary results. The results demonstrate that the approach presented in this work can provide economical and viable way of developing research reactor simulators

  4. Reactor Neutron Sources

    International Nuclear Information System (INIS)

    Aksenov, V.L.

    1994-01-01

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

  5. UCN sources at external beams of thermal neutrons. An example of PIK reactor

    International Nuclear Information System (INIS)

    Lychagin, E.V.; Mityukhlyaev, V.A.; Muzychka, A.Yu.; Nekhaev, G.V.; Nesvizhevsky, V.V.; Onegin, M.S.; Sharapov, E.I.; Strelkov, A.V.

    2016-01-01

    We consider ultracold neutron (UCN) sources based on a new method of UCN production in superfluid helium ("4He). The PIK reactor is chosen as a perspective example of application of this idea, which consists of installing "4He UCN source in the beam of thermal or cold neutrons and surrounding the source with moderator-reflector, which plays the role of cold neutron (CN) source feeding the UCN source. CN flux in the source can be several times larger than the incident flux, due to multiple neutron reflections from the moderator–reflector. We show that such a source at the PIK reactor would provide an order of magnitude larger density and production rate than an analogous source at the ILL reactor. We estimate parameters of "4He source with solid methane (CH_4) or/and liquid deuterium (D_2) moderator–reflector. We show that such a source with CH_4 moderator–reflector at the PIK reactor would provide the UCN density of ~1·10"5 cm"−"3, and the UCN production rate of ~2·10"7 s"−"1. These values are respectively 1000 and 20 times larger than those for the most intense UCN user source. The UCN density in a source with D_2 moderator-reflector would reach the value of ~2·10"5 cm"−"3, and the UCN production rate would be equal ~8·10"7 s"−"1. Installation of such a source in a beam of CNs would slightly increase the density and production rate.

  6. Development of source range measurement instrument in Xi'an pulsed reactor

    CERN Document Server

    Wang Li

    2002-01-01

    Source range measurement instrument in Xi'an pulsed reactor is key equipment of low-side measuring in source range. At the same time, it is also weighty component of out-of-pile neutron-flux level observation system. The authors have done some researching and renovating based on the similar type devices used in nuclear reactor to improve the meter sensitivity, measuring range, noise proof features, reliability in running and maintainability which belong to the main performance index of the instrument. The design ideas, configurations, working principle, performance indexes, technique features and effect in utilizing are introduced briefly

  7. An ultracold neutron source at the NC State University PULSTAR reactor

    Science.gov (United States)

    Korobkina, E.; Wehring, B. W.; Hawari, A. I.; Young, A. R.; Huffman, P. R.; Golub, R.; Xu, Y.; Palmquist, G.

    2007-08-01

    Research and development is being completed for an ultracold neutron (UCN) source to be installed at the PULSTAR reactor on the campus of North Carolina State University (NCSU). The objective is to establish a university-based UCN facility with sufficient UCN intensity to allow world-class fundamental and applied research with UCN. To maximize the UCN yield, a solid ortho-D 2 converter will be implemented coupled to two moderators, D 2O at room temperature, to thermalize reactor neutrons, and solid CH 4, to moderate the thermal neutrons to cold-neutron energies. The source assembly will be located in a tank of D 2O in the space previously occupied by the thermal column of the PULSTAR reactor. Neutrons leaving a bare face of the reactor core enter the D 2O tank through a 45×45 cm cross-sectional area void between the reactor core and the D 2O tank. Liquid He will cool the disk-shaped UCN converter to below 5 K. Independently, He gas will cool the cup-shaped CH 4 cold-neutron moderator to an optimum temperature between 20 and 40 K. The UCN will be transported from the converter to experiments by a guide with an inside diameter of 16 cm. Research areas being considered for the PULSTAR UCN source include time-reversal violation in neutron beta decay, neutron lifetime determination, support measurements for a neutron electric-dipole-moment search, and nanoscience applications.

  8. The Advanced Neutron Source (ANS) project: A world-class research reactor facility

    International Nuclear Information System (INIS)

    Thompson, P.B.; Meek, W.E.

    1993-01-01

    This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5x10 19 m -2 ·sec -1 . Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities

  9. Some preliminary design considerations for the ANS [Advanced Neutron Source] reactor cold source

    International Nuclear Information System (INIS)

    Henderson, D.L.

    1988-01-01

    Two areas concerned with the design of the Advanced Neutron Source (ANS) cold source have been investigated by simple one-dimensional calculations. The gain factors computed for a possible liquid nitrogen-15 cold source moderator are considerably below those computed for the much colder liquid deuterium moderator, as is reasonable considering the difference in moderator temperature. Nevertheless, nitrogen-15 does represent a viable option should safety related issues prohibit the use of deuterium as a moderating material. The slab geometry calculations have indicated that reflection of neutrons may be the dominant moderating mechanism and should be a consideration in the design of the cold source. 9 refs., 2 figs

  10. An improvement of source-jerk method for measuring high anti reactivities of reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Bosevski, T; Spiric, V [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)

    1966-07-01

    In this paper we modified the well known source jerk method (1) thus obtaining a method for experimental determination of negative reactivities of reactor systems by which, based on the basic idea of the source jerk method, a new experimental procedure and an exact analysis were developed. The analysis and numerical preparation allows direct application of the method to heavy water and graphite systems. Compared with the source jerk method the experimental procedure and the interpretation of results is faster, simpler and more exact (author)

  11. An improvement of source-jerk method for measuring high antireactivities of reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Bosevski, T; Spiric, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-10-15

    In this paper we modified the well known source jerk method /1/ thus obtaining a method for experimental determination of negative reactivities of reactor systems by which, based on the basic idea of the source jerk method, a new experimental procedure and an analysis were developed. The analysis and numerical preparation allows direct application of the method to heavy water and graphite systems. Compared with the source jerk method the experimental procedure and the interpretation of results is faster, simpler and more exact (author)

  12. An investigation of the closure problem applied to reactor accident source terms

    International Nuclear Information System (INIS)

    Brearley, I.R.; Nixon, W.; Hayns, M.R.

    1987-01-01

    The closure problem, as considered here, focuses attention on the question of when in current research programmes enough has been learned about the source terms for reactor accident releases. Noting that current research is tending to reduce the estimated magnitude of the aerosol component of atmospheric, accidental releases, several possible criteria for closure are suggested. Moreover, using the reactor accident consequence model CRACUK, the effect of gradually reducing the aerosol release fractions of a pressurized water reactor (PWR2) source term (as defined in the WASH-1400 study) is investigated and the implications of applying the suggested criteria to current source term research discussed. (author)

  13. An initial assessment of the Chernobyl-4 reactor accident release source

    International Nuclear Information System (INIS)

    Macdonald, H.F.; ApSimon, H.M.; Wilson, J.J.N.

    1986-07-01

    The long-range atmospheric dispersion model MESOS has been used to provide a preliminary evaluation of the effects over Western Europe of radioactivity released during the accident which occurred at the Chernobyl-4 reactor in the USSR in April 1986. The results of this analysis have been compared with observations during the first week or so following the accident of airborne contamination levels at a range of locations across Europe in order to obtain an estimate of accident release source. The work presented here was performed during the 6-8 weeks following the accident and the results obtained will be subject to refinement as more detailed data become available. However, at this early stage they indicate a release source for the Chernobyl accident, expressed as a fraction of the estimated reactor core inventory, of approx. 15-20% of the iodine and caesium isotopes, approx. 1% of the ruthenium and lesser amounts of the other fission products and actinides, together with an implied major fraction of the krypton and xenon noble gases. (author)

  14. Power Burst Reactor Facility as an epithermal neutron source for brain cancer therapy

    International Nuclear Information System (INIS)

    Wheeler, F.J.

    1986-01-01

    The Power Burst Facility (PBF) reactor is considered for modification to provide an intense, clean source of intermediate-energy (epithermal) neutrons desirable for clinical studies of neutron capture therapy (NCT) for malignant tumors. The modifications include partial replacement of the reflector, installation of a neutron-moderating, shifting region, additional shielding, and penetration of the present concrete shield with a collimating (and optionally) filtering region. The studies have indicated that the reactor, after these modifications, will be safely operable at full power (28 MW) within the acceptable limits of the plant protection systems. The neutron beam exiting from the collimator port is predicted to be of sufficient intensity (approx.10 10 neutrons/cm 2 -s) to provide therapeutic doses in very short irradiation times. The beam would be relatively free of undesirable fast neutrons, thermal neutrons and gamma rays. The calculated neutron energy spectrum and associated gamma rays in the beam were provided as input in simulation studies that used a computer model of a patient with a brain tumor to determine predicted dose rates to the tumor and healthy tissue. The results of this conceptual study indicate an intense, clean beam of epithermal neutrons for NCT clinical trials is attainable in the PBF facility with properly engineered design modifications. 9 refs., 11 figs., 3 tabs

  15. Determination of Source Term for an Annual Stack Release of Gas Reactor G.A. Siwabessy

    International Nuclear Information System (INIS)

    Sudiyati; Syahrir; Unggul Hartoyo; Nugraha Luhur

    2008-01-01

    Releases of radionuclide from the reactor are noble gases, halogenides and particulates. The measurements were carried out directly on the air monitoring system of the stack. The results of these measurements are compared with the annual Source-Term data from the Safety Analyses report (SAR) of RSG-GAS. The measurement results are smaller than the data reported in SAR document. (author)

  16. An advanced liquid hydrogen cold source for the NIST research reactor

    International Nuclear Information System (INIS)

    Williams, R.E.; Kopetka, P.; Rowe, J.M.

    1999-01-01

    A second-generation liquid hydrogen cold neutron source is currently being fabricated and will be installed in the NIST reactor early next year. The existing source has operated very successfully over the last four years, providing a six-fold increase in the cold neutron yield compared to the previous heavy ice source. The design of the new source is based on our operating experience with the existing LH 2 source and extensive neutron transport calculations using improved MCNP modeling and computational capabilities. Enhanced mechanical design and manufacturing tools are exploited in the fabrication of the advanced source, which is expected to nearly double the yield of the existing LH 2 source. (author)

  17. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  18. Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

    International Nuclear Information System (INIS)

    Kim, S.S.; Levine, S.H.

    1985-01-01

    An analytical/experimental method has been developed to monitor the subcritical reactivity and unfold the k/sub infinity/ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a Cf-252 neutron source placed sequentially in multiple positions in the core. Therefore, it is called the Asymmetric Multiple Position Neutron Source (AMPNS) method. The AMPNS method employs nucleonic codes to analyze the neutron multiplication of a Cf-252 neutron source. An optimization program, GPM, is utilized to unfold the k/sub infinity/ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn State Breazeale TRIGA Reactor (PSBR). A significant result of this study is that it provides a method to monitor the criticality of a damaged core during the recovery period

  19. From reactors to long pulse sources

    International Nuclear Information System (INIS)

    Mezei, F.

    1995-01-01

    We will show, that by using an adapted instrumentation concept, the performance of a continuous source can be emulated by one switch on in long pulses for only about 10% of the total time. This 10 fold gain in neutron economy opens up the way for building reactor like sources with an order of magnitude higher flux than the present technological limits. Linac accelerator driven spallation lends itself favorably for the realization of this kind of long pulse sources, which will be complementary to short pulse spallation sources, the same way continuous reactor sources are

  20. Measurement of reactivity in ADS reactors considering an exponential decay after an interruption in the external proton source

    Energy Technology Data Exchange (ETDEWEB)

    Henrice Junior, Edson; Gonçalves, Alessandro C., E-mail: ejunior@con.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro - RJ (Brazil)

    2017-07-01

    The online monitoring of reactivity in ADS reactors is of paramount importance for the operations of such systems. This work is dedicated to the prediction of reactivity from a decay of the neutron population after a pulse from the external source. For that, a pulse from an external source in an ADS reactor was simulated with Serpent Reactor Physics code. From the data obtained, it was possible to make an adjustment, based on a combination of exponentials. The coefficient of the exponential for the dominating term of the sum of exponentials is compared to the simplified solution of the neutron diffusion equation, thus obtaining the reactivity. The method used for the adjustment has the advantage of not requiring data that is equally spaced, and of being easily programmable, waiving the use of specific software for linear adjustments. The preliminary results of the research showed a 750 pcm deviation in relation to the value found of -3,630 pcm obtained through point kinetics, and as a result should be the object of further study. (author)

  1. Measurement of reactivity in ADS reactors considering an exponential decay after an interruption in the external proton source

    International Nuclear Information System (INIS)

    Henrice Junior, Edson; Gonçalves, Alessandro C.

    2017-01-01

    The online monitoring of reactivity in ADS reactors is of paramount importance for the operations of such systems. This work is dedicated to the prediction of reactivity from a decay of the neutron population after a pulse from the external source. For that, a pulse from an external source in an ADS reactor was simulated with Serpent Reactor Physics code. From the data obtained, it was possible to make an adjustment, based on a combination of exponentials. The coefficient of the exponential for the dominating term of the sum of exponentials is compared to the simplified solution of the neutron diffusion equation, thus obtaining the reactivity. The method used for the adjustment has the advantage of not requiring data that is equally spaced, and of being easily programmable, waiving the use of specific software for linear adjustments. The preliminary results of the research showed a 750 pcm deviation in relation to the value found of -3,630 pcm obtained through point kinetics, and as a result should be the object of further study. (author)

  2. Neutron source for a reactor

    International Nuclear Information System (INIS)

    Kobayashi, Hiromasa.

    1975-01-01

    Object: To easily increase a start-up power of a reactor without irradiation in other reactors. Structure: A neutron source comprises Cf 252 , a natural antimony rod, a layer of beryllium, and a vessel of neutron source. On upper and lower portion of Cf 252 are arranged natural antimony rods, which are surrounded by the Be layer, the entirety being charged into the vessel. The Cf 252 may emit neutron, has a half life more than a period of operating cycle of the reactor and is less deteriorated even irradiated by radioactive rays while being left within the reactor. The natural antimony rod is radioactivated by neutron from Cf 252 and neutron as reactor power increases to emit γ rays. The Be absorbs γ rays to emit the neutron. The antimony rod is irradiated within the reactor. Further, since the Cf 252 is small in neutron absorption cross section, it is hard to be deteriorated even while being inserted within the reactor. (Kamimura, M.)

  3. Advanced CANDU reactor: an optimized energy source of oil sands application

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Bock, D.; Miller, A.; Kuran, S.; Keil, H.; Fiorino, L.; Duffey, R.; Dunbar, R.B.

    2003-01-01

    Atomic Energy of Canada Limited (AECL) is developing the ACR-700 TM (Advanced CANDU Reactor-700 TM ) to meet customer needs for reduced capital cost, shorter construction schedule, high capacity factor while retaining the benefits of the CANDU experience base. The ACR-700 is based on the concept of CANDU horizontal fuel channels surrounded by heavy water moderator. The major innovation of this design is the use of slightly enriched uranium fuel in a CANFLEX bundle that is cooled by light water. This ensures: higher main steam pressures and temperatures providing higher thermal efficiency; a compact and simpler reactor design with reduced capital costs and shorter construction schedules; and reduced heavy water inventory compared to existing CANDU reactors. ACR-700 is not only a technically advanced and cost effective solution for electricity generating utilities, but also a low-cost, long-life and sustainable steam source for increasing Alberta's Oil Sand production rates. Currently practiced commercial surface mining and extraction of Oil Sand resources has been well established over the last three decades. But a majority of the available resources are somewhat deeper underground require in-situ extraction. Economic removal of such underground resources is now possible through the Steam Assisted Gravity Drainage (SAGD) process developed and proto-type tested in-site. SAGD requires the injection of large quantities of high-pressure steam into horizontal wells to form reduced viscosity bitumen and condensate mixture that is then collected at the surface. This paper describes joint AECL studies with CERI (Canadian Energy Research Institute) for the ACR, supplying both electricity and medium-pressure steam to an oil sands facility. The extensive oil sands deposits in northern Alberta are a very large energy resource. Currently, 30% of Canda's oil production is from the oil sands and this is expected to expand greatly over the coming decade. The bitumen deposits in the

  4. Solar energy as an alternate energy source to mixed oxide fuels in light-water cooled reactors

    International Nuclear Information System (INIS)

    Bertini, H.W.

    1977-01-01

    Supplemental information pertaining to the generic environmental impact statement on the Pu recycling process for mixed oxide light-water cooled reactors (GESMO) was requested from several sources. In particular, the role of alternate sources of energy was to be explored and the implications of these alternate sources to the question of Pu recycle in LWRs were to be investigated. In this vein, solar energy as an alternate source is the main subject of this report, along with other information related to solar energy. The general conclusion is that solar energy should have little effect on the decisions concerning GESMO

  5. Home brew technetium : clinical scale desktop plasma fusion neutron source to produce Tc99m as an alternative to industrial scale fission reactor sources

    International Nuclear Information System (INIS)

    Bosi, S.G.; Khachan, J.; Oborn, B.M.

    2011-01-01

    Full text: Tc-99m (decay product of Mo-99) accounts for ∼ 90% of world's production of radiopharmaceuticals. Recent unexpected shutdowns of two fission reactors and routine maintenance closures .e created a global shortage of Tc-99m, hence the large global effort to find alternative sources. This project aims to design and produce a novel prototype Mo-99/Tc-99m source. An operational desktop neutron source is available at the University of Sydney, employing a deuterium fusion-plasma to create 2.45 MeV neutrons. These neutrons will be used to activate Mo-98 thin an activation vessel. In one embodiment, the activation vessel contains an aqueous slurry or gel containing Mo-98 which converts to 0-99 upon activation. The decay product Tc-99m could then be milked, similar to existing Tc-99m generators. Monte Carlo will be :ed to assess yield versus size and geometry for various vessel designs. The neutron source filled with deuterium operating at 250 W, produces 3 x 106 neutrons continuously. The neutron flux can be increased ∼ 100-fold if the fill gas is 50% tritium and by another ∼ 100-1000-fold by increasing the power. This is being designed for local use, perhaps on the scale f one or a few hospitals, so the yield would not need to be industrial ;ale as with fission reactor sources. This device is low cost <$300 K) compared with cyclotrons and fission reactors.

  6. Compact reactor/ORC power source

    International Nuclear Information System (INIS)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

  7. A compact reactor/ORC power source

    International Nuclear Information System (INIS)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

  8. Recent activities of the international Group on Research Reactors (IGORR) and of the Advanced Neutron Source (ANS)

    International Nuclear Information System (INIS)

    West, C.D.

    1992-01-01

    The International Group on Research Reactors (IGORR) was formed in 1990 to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. The Advanced Neutron Source Project expects to complete conceptual design in mid-1992. In the present design concept, the neutron source is a heavy-water-cooled, moderated, and reflected reactor of about 350 MW(f) power. (author)

  9. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

    1987-01-01

    Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory

  10. Nuclear reactors sources of contamination - RA reactor in Vinca as an example; Nuklearni reaktori kao izvori kontaminacije - Primer reaktora RA u Vinci -

    Energy Technology Data Exchange (ETDEWEB)

    Mitrovic, S; Djukic, Z; Dobrosavljevic, N [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-03-15

    This study contains an analysis of possible contamination caused by reactor accidents. In addition to description of maximum credible accident impacts on the environment it covers the prevention and procedures to be undertaken in case of accident. The RA reactor was taken as an example. Since the probability of maximum credible accident is very low minor accident that could cause fission products release were analyzed. Analysis of the spent fuel storage pool at the RA reactor is included as well.

  11. 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

  12. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

    1990-09-01

    Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

  13. Very high flux steady state reactor and accelerator based sources

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  14. Source driven breeding thermal power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Ben-Gurion Univ. of the Negev, Beersheba

    1978-03-01

    The feasibility of fusion devices operating in the semi-catalyzed deuterium (SCD) mode and of high energy proton accelerators to provide the neutron sources for driving subcritical breeding light water power reactors is assessed. The assessment is done by studying the energy balance of the resulting source driven light water reactors (SDLWR) and comparing it with the energy balance of the reference light water hybrid reactors (LWHR) driven by a D-T neutron source (DT-LWHR). The conditions the non-DT neutron sources should satisfy in order to make the SDLWR viable power reactors are identified. It is found that in order for a SCD-LWHR to have the same overall efficiency as a DT-LWHR, the fusion energy gain of the SCD device should be at least one half that the DT device. The efficienct of ADLWRs using uranium targets is comparable with that of DT-LWHRs having a fusion energy gain of unity. Advantages and disadvantages of the DT-LWHR, SCD-LWHR and ADLWR are discussed. (aurthor)

  15. Dynamic modeling of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    March-Leuba, J.; Ibn-Khayat, M.

    1990-01-01

    The purpose of this paper is to provide a summary description and some applications of a computer model that has been developed to simulate the dynamic behavior of the advanced neutron source (ANS) reactor. The ANS dynamic model is coded in the advanced continuous simulation language (ACSL), and it represents the reactor core, vessel, primary cooling system, and secondary cooling systems. The use of a simple dynamic model in the early stages of the reactor design has proven very valuable not only in the development of the control and plant protection system but also of components such as pumps and heat exchangers that are usually sized based on steady-state calculations

  16. Reactor cold neutron source facility, the first in Japan

    International Nuclear Information System (INIS)

    Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

    1986-01-01

    In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

  17. Source driven breeding thermal power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

    1978-03-01

    Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)

  18. TOF powder diffractometer on a reactor source

    International Nuclear Information System (INIS)

    Bleif, H.J.; Wechsler, D.; Mezei, F.

    1999-01-01

    Complete text of publication follows. The performance of time-of-flight (TOF) methods on Long Pulse Spallation Sources can be studied at a reactor source. For this purpose a prototype TOF monochromator instrument will be installed at the KFKI reactor in Budapest. The initial setup will be a powder diffractometer with a resolution of δd/d down to 2 x 10 -3 at a wavelength of 1 A. The instrument uses choppers to produce neutron pulses of down to 10 μs FWHM. The optimal neutron source for a chopper instrument is a Long Pulse Spallation Source, but even on a continuous source simulations have shown that this instrument outperforms a conventional crystal monochromator powder diffractometer at high resolution. The main components of the TOF instrument are one double chopper defining the time resolution and two single choppers to select the wavelength range and to prevent frame overlap. For inelastic experiments a further chopper can be added in front of the sample. The neutron guide has a super-mirror coating and a curvature of 3500m. The total flight path is 20m and there are 24 single detectors in backscattering geometry. (author)

  19. Preliminary design concepts for the advanced neutron source reactor systems

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1988-01-01

    This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor

  20. Commissioning of the Opal reactor cold neutron source

    International Nuclear Information System (INIS)

    Thiering, R.; Lu, W.; Ullah, R.

    2006-01-01

    Full text: At OPAL, Australia's first cold neutron facility will form an essential part of the reactor's research programs. Fast neutrons, born in the core of a reactor, interact with a cryogenic material, in this case liquid deuterium, to give them very low energies ( 1 0 m eV). A cold neutron flux of 1.4 1 0 E 1 4 n /cm 2/ s is expected, with a peak in the energy spectrum at 4.2m eV. The cold neutron source reached cryogenic conditions for the first time in late 2005. The cold neutron source operates with a sub-cooled liquid Deuterium moderator at 24 K. The moderator chamber, which contains the deuterium, has been constructed from AlMg 5. The thermosiphon and moderator chamber are cooled by helium gas, in a natural convection thermosiphon loop. The helium refrigeration system utilises the Brayton cycle, and is fully insulated within a high vacuum environment. Despite the proximity of the cold neutron source to the reactor core, it has been considered as effectively separate to the reactor system, due to the design of its special vacuum containment vessel. As OPAL is a multipurpose research reactor, used for beam research as well as radiopharmaceutical production and industrial irradiations, the cold neutron source has been designed with a stand-by mode, to maximise production. The stand-by mode is a warm operating mode using only gaseous deuterium at ambient temperatures (∼ 3 00 K ), allowing for continued reactor operations whilst parts of the cold source are unavailable or in maintenance. This is the first time such a stand-by feature has been incorporated into a cold source facility

  1. Sources of gamma radiation in a reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Roos, Matts

    1959-05-15

    In a thermal reactor the gamma ray sources of importance for shielding calculations and related aspects are 1) fission, 2) decay of fission products, 3) capture processes in fuel, poison and other materials, 4) inelastic scattering in the fuel and 5) decay of capture products. The energy release and the gamma ray spectra of these sources have been compiled or estimated from the latest information available, and the results are presented in a general way to permit application to any thermal reactor, fueled with a mixture of {sup 235}U and {sup 238}U. As an example the total spectrum and the spectrum of radiation escaping from a fuel rod in the Swedish R3-reactor are presented.

  2. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  3. Proposal on the accelerator driven molten-salt reactor (ATW concept) benchmark calculations. (STAGE 1 - without an external neutron source)

    International Nuclear Information System (INIS)

    Svarny, J.; Mikolas, P.

    1999-01-01

    The first stage of ATW neutronic benchmark (without an external source), based on the simple modelling of two component concept is presented. The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark is not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (author)

  4. Fukushima - calculation of the reactor core inventory and storage pools Dai-ichi 1 to Dai-ichi 4, an estimation of a source term

    International Nuclear Information System (INIS)

    Krpelanova, M.; Carny, P.

    2011-01-01

    Inventory of the reactor core and spent fuel storage pool of the reactors at Dai-ichi 1 to Dai-ichi 4 was determined to need a realistic estimate of the source (released into the atmosphere environment) and modelling of radiological impact of the events in Fukushima NPP. Calculations of inventories were carried out by the methodology that is used in systems to support emergency response and crisis management anymore. Calculations were made based on a model that respects knowledge of real fuels and fuel cycles for individual reactors Dai-ichi. Necessary input data for training the model and calculate inventories are obtained from the IAEA PRIS database.

  5. Trench reactor: an overview

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

    1988-01-01

    Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

  6. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

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

  7. Fusion reactors as a future energy source

    International Nuclear Information System (INIS)

    Seifritz, W.

    A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

  8. Characterization of noise sources in nuclear power reactors

    International Nuclear Information System (INIS)

    Andhill, Gustav

    2004-03-01

    Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D

  9. Characterization of noise sources in nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Andhill, Gustav

    2004-03-01

    Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D.

  10. Fast reactor physics - an overview

    International Nuclear Information System (INIS)

    Lee, S.M.

    2004-01-01

    An introduction to the basic features of fast neutron reactors is made, highlighting the differences from the more conventional thermal neutron reactors. A discussion of important feedback reactivity mechanisms is given. Then an overview is presented of the methods of fast reactor physics, which play an important role in the successful design and operation of fast reactors. The methods are based on three main elements, namely (i) nuclear data bases, (ii) numerical methods and computer codes, and (iii) critical experiments. These elements are reviewed and the present status and future trends are summarized. (author)

  11. Research reactor of the future: The advanced neutron source

    International Nuclear Information System (INIS)

    Appleton, B.; West, C.

    1994-01-01

    Agents for cancer detection and treatment, stronger materials, better electronic gadgets, and other consumer and industrial products - these are assured benefits of a research reactor project proposed for Oak Ridge. Just as American companies have again assumed world leadership in producing semiconductor chips as well as cars and trucks, the United States is poised to retake the lead in neutron science by building and operating the $2.9 billion Advanced Neutron Source (ANS) research reactor by the start of the next century. In 1985, the neutron community, led by ORNL researchers, proposed a pioneering project, later called the ANS. Scheduled to begin operation in 2003, the ANS is seen not only as a replacement for the aging HFIR and HFBR but also as the best laboratory in the world for conducting neutron-based research

  12. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    International Nuclear Information System (INIS)

    Hultqvist, Pontus

    2004-09-01

    In this thesis the possibility of locating vibrating noise sources in a nuclear reactor core from the neutron noise has been investigated using different localization methods. The influence of the vibrating noise source has been considered to be a small perturbation of the neutron flux inside the reactor. Linear perturbation theory has been used to construct the theoretical framework upon which the localization methods are based. Two different cases have been considered: one where a one-dimensional one-group model has been used and another where a two-dimensional two-energy group noise simulator has been used. In the first case only one localization method is able to determine the position with good accuracy. This localization method is based on finding roots of an equation and is sensitive to other perturbations of the neutron flux. It will therefore work better with the assistance of approximative methods that reconstruct the noise source to determine if the results are reliable or not. In the two-dimensional case the results are more promising. There are several different localization techniques that reproduce both the vibrating noise source position and the direction of vibration with enough precision. The approximate methods that reconstruct the noise source are substantially better and are able to support the root finding method in a more constructive way. By combining the methods, the results will be more reliable

  13. Productivity of a nuclear chemical reactor with gamma radioisotopic sources

    International Nuclear Information System (INIS)

    Anguis T, C.

    1975-01-01

    According to an established mathematical model of successive Compton interaction processes the made calculations for major distances are extended checking the acceptability of the spheric geometry model for the experimental data for radioisotopic sources of Co-60 and Cs-137. Parameters such as the increasing factor and the absorbed dose served as comparative base. calculations for the case of a punctual source succession inside a determined volume cylinder are made to obtain the total dose, the deposited energy by each photons energetic group and the total absorbed energy inside the reactor. Varying adequately the height/radius relation for different cylinders, the distinct energy depositions are compared in each one of them once a time standardized toward a standard value of energy emitted by the reactor volume. A relation between the quantity of deposited energy in each point of the reactor and the conversion values of chemical species is established. They are induced by electromagnetic radiation and that are reported as ''G'' in the scientific literature (number of molecules formed or disappeared by each 100 e.v. of energy). Once obtained the molecular performance inside the reactor for each type of geometry, it is optimized the height/radius relation according to the maximum production of molecules by unity of time. It is completed a bibliographical review of ''G'' values reported by different types of aqueous solutions with the purpose to determine the maximum performance of molecular hydrogen as a function of pH of the solution and of the used type of solute among other factors. Calculations for the ethyl bromide production as an example of one of the industrial processes which actually work using the gamma radiation as reactions inductor are realized. (Author)

  14. Encapsulated nuclear heat source reactors for energy security

    International Nuclear Information System (INIS)

    Greenspan, E.; Susplugas, A.; Hong, S.G.; Monti, L.; Sumini, M.; Okawa, T.

    2006-01-01

    A spectrum of Encapsulated Nuclear Heat Source (ENHS) reactors have been conceptually designed over the last few years; they span a power range from 10 MWe to -200 MWe and consider a number of coolants and fuel types. Common features of all these designs include very long life cores - exceeding 20 effective full power years; nearly zero burnup reactivity swing; natural circulation; superb safety; autonomous load following capability; simplicity of operation and maintenance. ENHS reactors could be of particular interest for providing electricity, thermal energy and, possibly, desalinated water to communities that are not connected to a central electricity grid such as to many pacific islands and to remote communities in the mainland of different countries. ENHS reactors provide energy security by virtue of a couple of features: (1) Once an ENHS reactor is commissioned, the community has assured clean energy supply for at least 20 years without needing fuel supply. (2) The energy value of the fuel loaded (in the factory) in the ENHS module is preserved; what is needed for generating energy for additional 20+ years is to remove the fission products, add depleted uranium for makeup fuel, refabricate fuel rods and load into a new module. This fuel recycling is envisioned done by either the supplier country or by a regional or international fuel cycle centre. As the ENHS module is replaced at its entirety at the end of the core life - that is brought about by radiation damage, the ENHS plant life is likely to last for over 100 years. The above features also offer exceptional stability in the price of energy generated by the ENHS reactor. The reference ENHS design will be described followed by a brief description of the design options developed and a summary of their performance characteristics

  15. Fabrication development for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Pace, B.W.; Copeland, G.L.

    1995-08-01

    This report presents the fuel fabrication development for the Advanced Neutron Source (ANS) reactor. The fuel element is similar to that successfully fabricated and used in the High Flux Isotope Reactor (HFIR) for many years, but there are two significant differences that require some development. The fuel compound is U 3 Si 2 rather than U 3 O 8 , and the fuel is graded in the axial as well as the radial direction. Both of these changes can be accomplished with a straightforward extension of the HFIR technology. The ANS also requires some improvements in inspection technology and somewhat more stringent acceptance criteria. Early indications were that the fuel fabrication and inspection technology would produce a reactor core meeting the requirements of the ANS for the low volume fraction loadings needed for the highly enriched uranium design (up to 1.7 Mg U/m 3 ). Near the end of the development work, higher volume fractions were fabricated that would be required for a lower- enrichment uranium core. Again, results look encouraging for loadings up to ∼3.5 Mg U/m 3 ; however, much less evaluation was done for the higher loadings

  16. Promising design options for the encapsulated nuclear heat source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Conway, L.; Carelli, M.D.; Dzodzo, M. [Westinghouse Science and Technology, Pittsburgh, PA (United States); Hossain, Q.; Brown, N.W. [Lawrence Livermore National Lab., CA (United States); Wade, D.C.; Sienick, J.J. [Argonne National Lab., IL (United States); Greenspan, E.; Kastenberg, W.E.; Saphier, D. [University of California Dept of Nuclear Engineering, Berkeley, CA (United States)

    2001-07-01

    Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

  17. Promising design options for the encapsulated nuclear heat source reactor

    International Nuclear Information System (INIS)

    Conway, L.; Carelli, M.D.; Dzodzo, M.; Hossain, Q.; Brown, N.W.; Wade, D.C.; Sienick, J.J.; Greenspan, E.; Kastenberg, W.E.; Saphier, D.

    2001-01-01

    Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

  18. 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)

  19. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Paulo Cleber Mendonca

    2002-12-01

    In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)

  20. Method of judging leak sources in a reactor container

    International Nuclear Information System (INIS)

    Maeda, Katsuji.

    1984-01-01

    Purpose: To enable exact judgement for leak sources upon leak accident in a reactor container of BWR type power plants as to whether the sources are present in the steam system or coolant system. Method: If leak is resulted from the main steam system, the hydrogen density in the reactor container is about 170 times as high as the same amount of leak from the reactor water. Accordingly, it can be judged whether the leak source is present in the steam system or reactor water system based on the change in the indication of hydrogen densitometer within the reactor container, and the indication from the drain amount from the sump in the container or the indication of a drain flow meter in the container dehumidifier. Further, I-131, Na-24 and the like as the radioactive nucleides in sump water of the container are measured to determine the density ratio R = (I-131)/(Na-24), and it is judged that the leak is resulted in nuclear water if the density ratio R is equal to that of reactor water and that the leak is resulted from the main steam or like other steam system if the density ratio R is higher than by about 100 times than that of reactor water. (Horiuchi, T.)

  1. Aerosol behavior and light water reactor source terms

    International Nuclear Information System (INIS)

    Abbey, F.; Schikarski, W.O.

    1988-01-01

    The major developments in nuclear aerosol modeling following the accident to pressurized water reactor Unit 2 at Three Mile Island are briefly reviewed and the state of the art summarized. The importance and implications of these developments for severe accident source terms for light water reactors are then discussed in general terms. The treatment is not aimed at identifying specific source term values but is intended rather to illustrate trends, to assess the adequacy of the understanding of major aspects of aerosol behavior for source term prediction, and demonstrate in qualitative terms the effect of various aspects of reactor design. Areas where improved understanding of aerosol behavior might lead to further reductions in current source terms predictions are also considered

  2. High Flux Isotope Reactor cold neutron source reference design concept

    International Nuclear Information System (INIS)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R.

    1998-05-01

    In February 1995, Oak Ridge National Laboratory's (ORNL's) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH 2 ) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH 2 cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept

  3. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  4. Advanced neutron source reactor probabilistic flow blockage assessment

    International Nuclear Information System (INIS)

    Ramsey, C.T.

    1995-08-01

    The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

  5. Source-to-incident flux relation for a tokamak fusion test reactor blanket module

    International Nuclear Information System (INIS)

    Imel, G.R.

    1982-01-01

    The source-to-incident 14-MeV flux relation for a blanket module on the Tokamak Fusion Test Reactor is derived. It is shown that assumptions can be made that allow an analytical expression to be derived, using point kernel methods. In addition, the effect of a nonuniform source distribution is derived, again by relatively simple point kernel methods. It is thought that the methodology developed is valid for a variety of blanket modules on tokamak reactors

  6. Acceleration of relativistic electrons in plasma reactors and non-linear spectra of cosmic radio sources

    International Nuclear Information System (INIS)

    Kaplan, S.A.; Lomadze, R.D.

    1978-01-01

    A second approximation to the theory of turbulent plasma reactors in connection with the problem of interpretation of the non-linear spectra of cosmic radio sources has been investigated by the authors (Kaplan and Lomadze, 1977; Lomadze, 1977). The present paper discusses the basic results received for a Compton reactor with plasma waves of phase velocities smaller than the velocity of light, as well as for the synchrotron reactor. The distortion of the distribution function of relativistic electrons caused by their diffusion from the reactor is also presented as an example. (Auth.)

  7. Investigating The Integral Control Rod Worth Of The Miniature Neutron Source Reactor MNSR

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    Determining control rod characteristics is an essential problem of nuclear reactor analysis. In this research, the integral control rod worth of the miniature neutron source reactor MNSR is investigated. Some other parameters of the nuclear reactor, such as core excess reactivity, shut down margin, are also calculated. Group constants for all reactor components are generated by the WIMSD code and then are used in the CITATION code to solve the neutron diffusion equations. The maximum relative error of the calculated results compared with the measurement data is about 3.5%. (author)

  8. Revised reactor accident source terms in the U.S. and implementation for light water reactors

    International Nuclear Information System (INIS)

    Soffer, L.; Lee, J.Y.

    1992-01-01

    Current NRC reactor accident source terms used for licensing are contained in Regulatory Guides 1.3 and 1.4 and specify that 100 % of the core inventory of noble gases and 25 % of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental (I 2 ) iodine. These assumptions have strongly affected present nuclear plant designs. Severe accident research results have confirmed that although the current source term is very substantial and has resulted in a very high level of plant capability, the present source term is no longer compatible with a realistic understanding of severe accidents. The NRC has issued a proposed revision of the reactor accident source terms as part of several regulatory activities to incorporate severe accident insights for future plants. A revision to 10 CFR 100 is also being proposed to specify site criteria directly and to eliminate source terms and doses for site evaluation. Reactor source terms will continue to be important in evaluating plant designs. Although intended primarily for future plants, existing and evolutionary power plants may voluntarily apply revised accident source term insights as well in licensing. The proposed revised accident source terms are presented in terms of fission product composition, magnitude, timing and iodine chemical form. Some implications for light water reactors are discussed. (author)

  9. Integral Fast Reactor: A future source of nuclear energy

    International Nuclear Information System (INIS)

    Southon, R.

    1993-01-01

    Argonne National Laboratory is developing a reactor concept that would be an important part of the worlds energy future. This report discusses the Integral Fast Reactor (IFR) concept which provides significant improvements over current generation reactors in reactor safety, plant complexity, nuclear proliferation, and waste generation. Two major facilities, a reactor and a fuel cycle facility, make up the IFR concept. The reactor uses fast neutrons and metal fuel in a sodium coolant at atmospheric pressure that relies on laws of physics to keep it safe. The fuel cycle facility is a hot cell using remote handling techniques for fabricating reactor fuel. The fuel feed stock includes spent fuel from the reactor, and potentially, spent light water reactor fuel and plutonium from weapons. This paper discusses the unique features of the IFR concept and the differences the quality assurance program has from current commercial practices. The IFR concept provides an opportunity to design a quality assurance program that makes use of the best contemporary ideas on management and quality

  10. Reactor installation and maintenance for the Advanced Neutron Source

    International Nuclear Information System (INIS)

    Smith, B.R.

    1993-01-01

    Advanced Neutron Source (ANS) reactor assembly components have been modeled in great detail in IGRIP in order to realistically simulate preliminary installation and maintenance processes. Animation of these processes has been captured in a 15-minute video with narration. Approximately 90% of the parts were initially translated from CADAM (a two-dimensional drawing package) to IGRIP and then revolved or extruded. IGRIP's IGES translator greatly reduced the time required to perform this operation. The interfacing of devices in the work cell has identified numerous design inconsistencies. Most of the modeled reactor components are devices with a single degree of freedom (DOF) however, some of the slanted experiments required 6 DOF so that they could be removed at an angle in order to clear the reflector vessel flanges. IGRIP's collision detection feature proved to be extremely helpful in determining interferences when removing the experiments. The combination of three-dimensional visualization and collision detection allows engineers to clearly and easily visualize potential design problems before the construction phase of the project

  11. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (P fus ∼ 10-100 MW, β N ∼ 2-3, Q p ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

  12. High flux isotope reactor cold source preconceptual design study report

    International Nuclear Information System (INIS)

    Selby, D.L.; Bucholz, J.A.; Burnette, S.E.

    1995-12-01

    In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH 2 moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project

  13. Research reactor fuel - an update

    International Nuclear Information System (INIS)

    Finlay, M.R.; Ripley, M.I.

    2003-01-01

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  14. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  15. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  16. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    International Nuclear Information System (INIS)

    Rosenthal, Murray Wilford

    2009-01-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  17. Revised accident source terms for light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Soffer, L. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  18. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    Science.gov (United States)

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

  19. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  20. A multi-stage-flash desalination plant of relative small performance with an integrated pressurized water reactor as a nuclear heat source

    International Nuclear Information System (INIS)

    Peltzer, M.; Petersen, G.

    1976-01-01

    In the Krupp-GKSS joint study MINIPLEX the requirements for seawater-desalination-plants with a performance in the range of 10,000 to 80,000 m 3 /d heated by a nuclear reactor are investigated. The reactor concept is similar to the integrated pressurized water reactor (IPWR) of the nuclear ship OTTO HAHN. The calculated costs of the desalinated water show, that due to the fuel cost advantages of reactors small and medium nuclear desalination plants are economically competetive with oil-fired plants since the steep rise of oil price in autumn 1973. (orig.) [de

  1. A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2017-06-26

    A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

  2. Combustion flame-plasma hybrid reactor systems, and chemical reactant sources

    Science.gov (United States)

    Kong, Peter C

    2013-11-26

    Combustion flame-plasma hybrid reactor systems, chemical reactant sources, and related methods are disclosed. In one embodiment, a combustion flame-plasma hybrid reactor system comprising a reaction chamber, a combustion torch positioned to direct a flame into the reaction chamber, and one or more reactant feed assemblies configured to electrically energize at least one electrically conductive solid reactant structure to form a plasma and feed each electrically conductive solid reactant structure into the plasma to form at least one product is disclosed. In an additional embodiment, a chemical reactant source for a combustion flame-plasma hybrid reactor comprising an elongated electrically conductive reactant structure consisting essentially of at least one chemical reactant is disclosed. In further embodiments, methods of forming a chemical reactant source and methods of chemically converting at least one reactant into at least one product are disclosed.

  3. Cylindrical IEC neutron source design for driven research reactor operation

    International Nuclear Information System (INIS)

    Miley, G.H.; Ulmen, B.; Amadio, G.; Leon, H.; Hora, H.

    2009-01-01

    A resurgence in nuclear power use is now underway worldwide. However, due many university research reactors shutdown, they must rely on using subcritical assemblies which employs a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The source is inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory. (author)

  4. WWER type reactors used as multipurpose nuclear power sources

    International Nuclear Information System (INIS)

    Fiala, J.; Mulak, J.

    1976-01-01

    Safety aspects are assessed of the siting of nuclear power installations in the vicinity of large housing estates and in areas with a high population density, mainly the aspect of the liquidation of the consequences of the maximum credible accident, i.e., the transversal rupture of the primary coolant circuit. The application of WWER type reactors as multipurpose nuclear power sources in Czechoslovakia is justified. It is shown that such a multipurpose nuclear power source differs from a purely condensation nuclear power plant mainly in the design of the secondary stage. The possibilities of such projects are indicated with a view to power and heat operation. (F.M.)

  5. Pulsed TRIGA reactor as substitute for long pulse spallation neutron source

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1999-01-01

    TRIGA reactor cores have been used to demonstrate various pulsing applications. The TRIGA reactor fuel (U-ZrH x ) is very robust especially in pulsing applications. The features required to produce 50 pulses per second have been successfully demonstrated individually, including pulse tests with small diameter fuel rods. A partially optimized core has been evaluated for pulses at 50 Hz with peak pulsed power up to 100 MW and an average power up to 10 MW. Depending on the design, the full width at half power of the individual pulses can range between 2000 μsec to 3000 μsec. Until recently, the relatively long pulses (2000 μsec to 3000 μsec) from a pulsed thermal reactor or a long pulse spallation source (LPSS) have been considered unsuitable for time-of-flight measurements of neutron scattering. More recently considerable attention has been devoted to evaluating the performance of long pulse (1000 to 4000 μs) spallation sources for the same type of neutron measurements originally performed only with short pulses from spallation sources (SPSS). Adequate information is available to permit meaningful comparisons between CW, SPSS, and LPSS neutron sources. Except where extremely high resolution is required (fraction of a percent), which does require short pulses, it is demonstrated that the LPSS source with a 1000 msec or longer pulse length and a repetition rate of 50 to 60 Hz gives results comparable to those from the 60 MW ILL (CW) source. For many of these applications the shorter pulse is not necessarily a disadvantage, but it is not an advantage over the long pulse system. In one study, the conclusion is that a 5 MW 2000 μsec LPSS source improves the capability for structural biology studies of macromolecules by at least a factor of 5 over that achievable with a high flux reactor. Recent studies have identified the advantages and usefulness of long pulse neutron sources. It is evident that the multiple pulse TRIGA reactor can produce pulses comparable to

  6. Review of the accident source terms for aluminide fuel: Application to the BR2 reactor

    International Nuclear Information System (INIS)

    Joppen, F.

    2005-01-01

    A major safety review of the BR2, a material test reactor, is to be conducted for the year 2006. One of the subjects selected for the safety review is the definition of source terms for emergency planning and in particular the development of accident scenarios. For nuclear power plants the behaviour of fuel under accident conditions is a well studied object. In case of non-power reactors this basic knowledge is rather scarce. The usefulness of information from power plant fuels is limited due to the differences in fuel type, power level and thermohydraulical conditions. First investigation indicates that using data from power plant fuel leads to an overestimation of the source terms. Further research on this subject could be very useful for the research reactor community, in order to define more realistic source terms and to improve the emergency preparedness. (author)

  7. Efficiency factor of a chemical nuclear reactor with gamma sources

    International Nuclear Information System (INIS)

    Anguis T, C.

    1975-01-01

    A chemonuclear reactor is simulated in order to calculate the efficiency factor of molecular species in chemical reactions induced by gamma radiation, with the purpose to obtain information for its design and consider the electromagnetic energy as a possible solution to the present problem of energy. The research is based on a mathematical model of succesive Compton processes applied to spherical and cylindrical geometry and corroborated through the absorbed dose and the experimental date of the increase factor, for the radioisotopic sources Co-60 and Cs-137 relating the quantity of energy deposited into various cylinders with the G value, the relation radius/height of the reactor is optimized according to the molecular production. This is illustrated with the radiolysis of a solution of CH 3 OH/H 2 O which forms H 2 and with the obtainment of C 2 H 5 Br that represents and industrial process induced radioactively. The results show a greater energy deposition with Cs-137 but a larger production of H 2 /hr with Co-60, and besides we can find high production values of C 2 H 5 Br. The cylinder with more advantages is that whose relation R/H is of 0.5. It can be concluded that the final selection of the reactor should be made after a more intense study of the used isotope and the source activity. The efficiency factor of H 2 can be increased selecting the appropriate type and concentration of solute of the irradiated aqueous solutions

  8. Reactor Core Design and Analysis for a Micronuclear Power Source

    Directory of Open Access Journals (Sweden)

    Hao Sun

    2018-03-01

    Full Text Available Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

  9. Lowering the enrichment of the Syrian miniature neutron source reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.

    1999-01-01

    An investigation of lowering the fuel enrichment of MNSR was realized. A 3-D neutronic model was developed for the analysis of the reactor. It was found that lower number of fuel elements is needed when UO 2 is used with 5.45 g of 235 U content in each fuel rod. Sensitivity of the reactor to the purity of the beryllium reflector proved to be an important factor in determining the reactor neutronics as well as the weight of loaded fuel in the core. Inherent safety features of low excess reactivity and shutdown margins are preserved and enhanced. Average thermal fluxes over different zones of the core are kept very much unchanged

  10. Lowering the enrichment of the Syrian miniature neutron source reactor

    International Nuclear Information System (INIS)

    Kamis, I.; Khattab, K.

    1999-01-01

    An investigation of lowering the fuel enrichment of MNSR was realized. A 3-D neutronic model was developed for the analysis of the reactor. It was found that lower number of fuel elements is needed when UO 2 is used with 5.45 g of 235 U content in each fuel rod. sensitivity of the reactor to the purity of the beryllium reflector proved to be an important factor in determining the reactor neutronics as well as the weight of loaded fuel in the core. Inherent safety feature of low excess reactivity and shutdown margins are preserved and enhanced. average thermal fluxes over different zones of the core are kept very much unchanged. (author)

  11. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-01-01

    Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  12. Source term determination from subcritical multiplication measurements at Koral-1 reactor

    International Nuclear Information System (INIS)

    Blazquez, J.B.; Barrado, J.M.

    1978-01-01

    By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

  13. Subcriticality calculation in nuclear reactors with external neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br

    2007-07-01

    The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

  14. Subcriticality calculation in nuclear reactors with external neutron sources

    International Nuclear Information System (INIS)

    Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2007-01-01

    The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

  15. Small-angle neutron scattering at pulsed sources compared to reactor sources

    International Nuclear Information System (INIS)

    Hjelm, R.P. Jr.; Seeger, P.A.; Thiyagarajan, P.

    1990-01-01

    Detailed comparisons of measurements made on small-angle neutron scattering instruments at pulsed spallation and reactor sources show that the results from the two types of instruments are comparable. It is further demonstrated that spallation instruments are preferable for measurements in the mid-momentum transfer domain or when a large domain is needed. 8 refs., 2 figs

  16. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

    1995-08-01

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

  17. Cold neutron source conceptual designing for Tehran Research Reactor

    International Nuclear Information System (INIS)

    Khajvand, N.; Mirvakili, S.M.; Faghihi, F.

    2016-01-01

    Highlights: • Cold neutron source conceptual designing for Tehran research reactor is carried out. • Type and geometry of moderator and dimensions of cold neutron source are analyzed. • Liquid hydrogen with more ortho-concentration can be better option as moderator. - Abstract: A cold neutron source (CNS) conceptual designing for the Tehran Research Reactor (TRR) were carried out using MCNPX code. In this study, a horizontal beam tube of the core which has appropriate the highest thermal flux is selected and parametric analysis to choose the type and geometry of the moderator, and the required CNS dimensions for maximizing the cold neutron production was performed. In this design the moderator cell has a spherical annulus structure, and the cold neutron flux and its brightness are calculated together with the nuclear heat load of the CNS for a variety of materials including liquid hydrogen, liquid deuterium, and solid methane. Based on our study, liquid hydrogen with more ortho-concentration than para and solid methane are the best options.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Yoder, G.L. Jr.; Dixon, J.R.; Elkassabgi, Y.; Felde, D.K.; Giles, G.E.; Harrington, R.M.; Morris, D.G.; Nelson, W.R.; Ruggles, A.E.; Siman-Tov, M.; Stovall, T.K.

    1994-05-01

    The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

  20. 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.)

  1. Optimizing a three-element core design for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    West, C.D.

    1995-01-01

    Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ''no new inventions'' rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable 235 U has to be diluted with 238 U to reduce the enrichment. This paper describes the design and optimization of that three-element core

  2. Plant protection system optimization studies to mitigate consequences of large breaks in the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khayat, M.I.; March-Leuba, J.

    1993-01-01

    This paper documents some of the optimization studies performed to maximize the performance of the engineered safety features and scram systems to mitigate the consequences of large breaks in the primary cooling system of the Advanced Neutron Source (ANS) Reactor

  3. Consideration of emergency source terms for pebble-bed high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Tao, Liu; Jun, Zhao; Jiejuan, Tong; Jianzhu, Cao

    2009-01-01

    Being the last barrier in the nuclear power plant defense-in-depth strategy, emergency planning (EP) is an integrated project. One of the key elements in this process is emergency source terms selection. Emergency Source terms for light water reactor (LWR) nuclear power plant (NPP) have been introduced in many technical documents, and advanced NPP emergency planning is attracting attention recently. Commercial practices of advanced NPP are undergoing in the world, pebble-bed high-temperature gas-cooled reactor (HTGR) power plant is under construction in China which is considered as a representative of advanced NPP. The paper tries to find some pieces of suggestion from our investigation. The discussion of advanced NPP EP will be summarized first, and then the characteristics of pebble-bed HTGR relating to EP will be described. Finally, PSA insights on emergency source terms selection and current pebble-bed HTGR emergency source terms suggestions are proposed

  4. A hybrid source-driven method to compute fast neutron fluence in reactor pressure vessel - 017

    International Nuclear Information System (INIS)

    Ren-Tai, Chiang

    2010-01-01

    A hybrid source-driven method is developed to compute fast neutron fluence with neutron energy greater than 1 MeV in nuclear reactor pressure vessel (RPV). The method determines neutron flux by solving a steady-state neutron transport equation with hybrid neutron sources composed of peripheral fixed fission neutron sources and interior chain-reacted fission neutron sources. The relative rod-by-rod power distribution of the peripheral assemblies in a nuclear reactor obtained from reactor core depletion calculations and subsequent rod-by-rod power reconstruction is employed as the relative rod-by-rod fixed fission neutron source distribution. All fissionable nuclides other than U-238 (such as U-234, U-235, U-236, Pu-239 etc) are replaced with U-238 to avoid counting the fission contribution twice and to preserve fast neutron attenuation for heavy nuclides in the peripheral assemblies. An example is provided to show the feasibility of the method. Since the interior fuels only have a marginal impact on RPV fluence results due to rapid attenuation of interior fast fission neutrons, a generic set or one of several generic sets of interior fuels can be used as the driver and only the neutron sources in the peripheral assemblies will be changed in subsequent hybrid source-driven fluence calculations. Consequently, this hybrid source-driven method can simplify and reduce cost for fast neutron fluence computations. This newly developed hybrid source-driven method should be a useful and simplified tool for computing fast neutron fluence at selected locations of interest in RPV of contemporary nuclear power reactors. (authors)

  5. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The law aims to perform regulations on enterprises of refining, processing and reprocessing of nuclear source and fuel materials and on establishment and operation of reactors to realize the peaceful and deliberate utilization of atomic energy according to the principle of the atomic energy basic law. Regulations of use of internationally regulated substances are also envisaged to observe international agreements. Basic concepts and terms are defined, such as: atomic energy; nuclear fuel material; nuclear source material; reactor; refining; processing; reprocessing and internationally regulated substance. Any person besides the Power Reactor and Nuclear Fuel Material Developing Corporation who undertakes refining shall be designated by the Prime Minister and the Minister of International Trade and Industry. An application shall be filed to the ministers concerned, listing name and address of the person, name and location of the refining works, equipment and method of refining, etc. The permission of the Prime Minister is necessary for any person who engages in processing. An application shall be filed to the Prime Minister, listing name and address of the person, name and location of the processing works and equipment and method of processing, etc. Permission of the Prime Minister, the Minister of International Trade and Industry or the Minister of Transport is necessary for any person who sets up reactors. An application shall be filed to the minister concerned, listing name and address of the person, purpose of operation, style, thermal output of reactor and number of units, etc. (Okada, K.)

  6. A small-scale modular reactor for electric source for remote places

    International Nuclear Information System (INIS)

    Anon.

    2002-01-01

    Use of a small-scale modular reactor (SMR) as an electric source for remote places is one of scenarios for actual use of SMR parallel to alternative source of present nuclear power stations and co-generation source at urban suburbs, there is not only an actual experience to construct and operate for power source for military use in U.S.A. on 1950s to 1960s, but also four nuclear reactors (LWGR, 12 MW) in Vilyvino Nuclear Power Station in far northern district in Russia are under operation. Recently, Department of Energy in U.S.A. prepared the 'Report to Congress on Small Modular Nuclear Reactors' evaluating on feasibility of SMR as a power source for remote places according to requirement of the Congress. This report evaluated a feasibility study on nine SMRs in the world with 10 to 50 MW of output as electric source for remote places on economical efficiency and so on, together with analysis of their design concepts, to conclude that 'they could perform beginning of operations on 2000s because of no large technical problems and keeping a level capable of competing with power generation cost at remote place on its present economical efficiency'. Here was introduced on outlines of this report. (G.K.)

  7. Inverse kinetics method with source term for subcriticality measurements during criticality approach in the IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Loureiro, Cesar Augusto Domingues; Santos, Adimir dos

    2009-01-01

    In reactor physics tests which are performed at the startup after refueling the commercial PWRs, it is important to monitor subcriticality continuously during criticality approach. Reactivity measurements by the inverse kinetics method are widely used during the operation of a nuclear reactor and it is possible to perform an online reactivity measurement based on the point reactor kinetics equations. This technique is successful applied at sufficiently high power level or to a core without an external neutron source where the neutron source term in point reactor kinetics equations may be neglected. For operation at low power levels, the contribution of the neutron source must be taken into account and this implies the knowledge of a quantity proportional to the source strength, and then it should be determined. Some experiments have been performed in the IPEN/MB-01 Research Reactor for the determination of the Source Term, using the Least Square Inverse Kinetics Method (LSIKM). A digital reactivity meter which neglects the source term is used to calculate the reactivity and then the source term can be determined by the LSIKM. After determining the source term, its value can be added to the algorithm and the reactivity can be determined again, considering the source term. The new digital reactivity meter can be used now to monitor reactivity during the criticality approach and the measured value for the reactivity is more precise than the meter which neglects the source term. (author)

  8. Scope of activities and organization of an interuniversity reactor institute

    International Nuclear Information System (INIS)

    de Bruin, M.

    1990-01-01

    The Reactor Instituut Delft was founded in 1958 and was at that time part of the Delft University of Technology. In 1969, the institute was converted into an interuniversity institute, owned and directed by the combined Dutch universities. Since 1987, the institute has again constituted part of the Delft University of Technology, still continuing its role as an interuniversity institute and with provisions in the organizational structure to secure this role. The major facility of the institute is the Hoger Onderwijs reactor, a 2-MW swimming pool reactor operated 24 h/day, 5 days/week. The reactor is used in neutron beam studies, reactor physics research, and for isotope production, neutron activation analysis (NAA), and commercial irradiation. The institute's 3-MeV Van de Graaff electron accelerator is mainly used for radiation chemistry. It can deliver subnanosecond high-current pulses and is provided with fast optical and conductivity measuring equipment. The variable energy positron source is being used for the study of defects at metal surfaces and interfaces. The experience obtained with this source is used in the development of a much stronger source as the basis of a positron microbeam in one of the reactor beam tubes

  9. The Chernobyl reactor accident source term: development of a consensus view

    International Nuclear Information System (INIS)

    Devell, L.; Guntay, S.; Powers, D.A.

    1995-11-01

    Ten years after the reactor accident at Chernobyl, a great deal more data is available concerning the events, phenomena, and processes that took place. The purpose of this document is to examine what is known about the radioactive materials released during the accident, a task that is substantially more difficult than it might first appear to be. The Chernobyl station, like other nuclear power plants, was not instrumented to characterize a disastrous accident. The accident was peculiar in the sense that radioactive materials were released, at least initially, in an exceptionally energetic plume and were transported far from the reactor site. Release of radioactivity from the plant continued for several days. Characterization of the contamination caused by the releases of radioactivity has had a much lower priority than remediation of the contamination. Consequently, an assessment of the Chernobyl accident source term must rely to a significant extent on inferential evidence. The assessment presented here begins with an examination of the core inventories of radioactive materials. In subsequent sections of the report, the magnitude and timing of the releases of radioactivity are described. Then, the composition, chemical forms, and physical forms of the releases are discussed. A number of more recent publications and results from scientists in Russia and elsewhere have significantly improved the understanding of the Chernobyl source term. Because of the special features of the reactor design and the peculiarities of the Chernobyl accident, the source term for the Chernobyl accident is of limited applicability to the safety analysis of other types of reactors

  10. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  11. PRISM reactor. An option for plutonium disposition?

    Energy Technology Data Exchange (ETDEWEB)

    Fehlinger, Sebastian; Friess, Friederike; Kuett, Moritz [IANUS, Technische Universitaet Darmstadt (Germany)

    2015-07-01

    The Power Reactor Innovative Small Module (PRISM) is sodium cooled fast reactor model. The energy output depends on the core configuration, however with an energy output of approximately 300 MWe, the PRISM reactor belongs to the class of small modular reactors. Beside using the reactor as a breeder reactor or for the transmutation of nuclear waste, it might also be used as a burner reactor for separated plutonium. This includes for example U.S.-American excess weapon-grade plutonium as well as separated reactor-grade plutonium. Recently, there has been an ongoing discussion in GB to use the PRISM reactor to dispose their excess civilian plutonium. Depending on the task, the core configuration varies slightly. We will present different layouts and the matching MCNP models, these models can then be used to conduct depletion calculations. From these results, analysis of the change in the plutonium isotopics in the spent fuel, the amount of fissioned plutonium, and the possible annual plutonium throughputs is possible.

  12. Advanced Neutron Source (ANS) Project

    International Nuclear Information System (INIS)

    Campbell, J.H.; Thompson, P.B.

    1994-01-01

    This report covers the progress made in 1993 in the following sections: (1) project management; (2) research and development; (3) design and (4) safety. The section on research and development covers the following: (1) reactor core development; (2) fuel development; (3) corrosion loop tests and analysis; (4) thermal-hydraulic loop tests; (5) reactor control and shutdown concepts; (6) critical and subcritical experiments; (7) material data, structure tests, and analysis; (8) cold source development; (9) beam tube, guide, and instrument development; (10) neutron transport and shielding; (11) I and C research and development; and (12) facility concepts

  13. Development of source term evaluation method for Korean Next Generation Reactor(III)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Geon Jae; Park, Jin Baek; Lee, Yeong Il; Song, Min Cheonl; Lee, Ho Jin [Korea Advanced Institue of Science and Technology, Taejon (Korea, Republic of)

    1998-06-15

    This project had investigated irradiation characteristics of MOX fuel method to predict nuclide concentration at primary and secondary coolant using a core containing 100% of all MOX fuel and development of source term evaluation tool. In this study, several prediction methods of source term are evaluated. Detailed contents of this project are : an evaluation of model for nuclear concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant using purely MOX fuel, suggestion of source term prediction method of NPP with a core using MOX fuel.

  14. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Energy Technology Data Exchange (ETDEWEB)

    Korobkina, E., E-mail: ekorobk@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Medlin, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Wehring, B.; Hawari, A.I. [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Huffman, P.R.; Young, A.R. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Beaumont, B. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Palmquist, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States)

    2014-12-11

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ∼40K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  15. Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term - Trial Calculation

    International Nuclear Information System (INIS)

    Grabaskas, David

    2016-01-01

    The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

  16. Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Denman, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Clark, Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Denning, Richard S. [Consultant, Columbus, OH (United States)

    2016-10-01

    The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

  17. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

  18. 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…

  19. A TRIGA reactor in an industrial laboratory

    International Nuclear Information System (INIS)

    Anders, Oswald U.

    1980-01-01

    The Dow TRIGA Reactor is a well established facility in its industrial environment. It is used extensively for internal Dow projects. The primary use of the TRIGA is as neutron source for NAA. It faces similar technical and organizational challenges as other TRIGA installations and over the years developed its own solutions

  20. The advanced neutron source - A world-class research reactor facility

    International Nuclear Information System (INIS)

    Thompson, P.B.; Meek, W.E.

    1993-01-01

    The advanced neutron source (ANS) is a new facility being designed at the Oak Ridge National Laboratory that is based on a heavy-water-moderated reactor and extensive experiment and user-support facilities. The primary purpose of the ANS is to provide world-class facilities for neutron scattering research, isotope production, and materials irradiation in the United States. The neutrons provided by the reactor will be thermalized to produce sources of hot, thermal, cold, very cold, and ultracold neutrons usable at the experiment stations. Beams of cold neutrons will be directed into a large guide hall using neutron guide technology, greatly enhancing the number of research stations possible in the project. Fundamental and nuclear physics, materials analysis, and other research pro- grams will share the neutron beam facilities. Sufficient laboratory and office space will be provided to create an effective user-oriented environment

  1. Design of an organic simplified nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

    2016-08-15

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  2. Design of an Organic Simplified Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-08-01

    Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  3. An accelerator based steady state neutron source

    International Nuclear Information System (INIS)

    Burke, R.J.; Johnson, D.L.

    1985-01-01

    Using high current, c.w. linear accelerator technology, a spallation neutron source can achieve much higher average intensities than existing or proposed pulsed spallation sources. With about 100 mA of 300 MeV protons or deuterons, the Accelerator Based Neutron Research Facility (ABNR) would initially achieve the 10 16 n/cm 2 .s thermal flux goal of the advanced steady state neutron source, and upgrading could provide higher steady state fluxes. The relatively low ion energy compared to other spallation sources has an important impact on R and D requirements as well as capital cost, for which a range of $300-450M is estimated by comparison to other accelerator-based neutron source facilities. The source is similar to a reactor source in most respects. It has some higher energy neutrons but fewer gamma rays, and the moderator region is free of many of the design constraints of a reactor, which helps to implement sources for various neutron energy spectra, many beam tubes, etc. With the development of multi-beam concept and the basis for currents greater than 100 mA that is assumed in the R and D plan, the ABNR would serve many additional uses, such as fusion materials development, production of proton-rich isotopes, and other energy and defense program needs

  4. The Chernobyl reactor accident source term: Development of a consensus view

    International Nuclear Information System (INIS)

    Guntay, S.; Powers, D.A.; Devell, L.

    1997-01-01

    In August 1986, scientists from the former Soviet Union provided the nuclear safety community with an impressively detailed account of what was then known about the Chernobyl accident. This included assessments of the magnitudes, rates, and compositions of radionuclide releases during the ten days following initiation of the accident. A summary report based on the Soviet report, the oral presentations, and the discussions with scientists from various countries was issued by the International Atomic Energy Agency shortly thereafter. Ten years have elapsed since the reactor accident at Chernobyl. A great deal more data is now available concerning the events, phenomena, and processes that took place. The purpose of this document is to examine what is known about the radioactive materials released during the accident. The accident was peculiar in the sense that radioactive materials were released, at least initially, in an exceptionally energetic plume and were transported far from the reactor site. Release of radioactivity from the plant continued for about ten days. A number of more recent publications and results from scientists in Russia and elsewhere have significantly improved our understanding of the Chernobyl source term. Because of the special features of the reactor design and the pecularities of the Chernobyl accident, the source term for the Chernobyl accident is of limited applicability of the safety analysis of other types of reactors

  5. Preparation and benchmarking of ANSL-V cross sections for advanced neutron source reactor studies

    International Nuclear Information System (INIS)

    Arwood, J.W.; Ford, W.E. III; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

    1987-01-01

    Research and development for the advanced neutron source (ANS) reactor is being funded by the US Dept. of Energy. This reactor is to provide the world's most intense steady-state source of low-energy neutrons for a national experimental user facility. Pseudo-problem-independent, multigroup cross-section libraries were generated to support ANS design work. The libraries, designated ANSL-V, are data bases in AMPX master format for subsequent generation of problem-dependent cross sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, and MORSE. Included in ANSL-V are 123-material P 3 neutron, 46-material P 0 or P 6 secondary gamma-ray production (SGRP), and 34-material P 6 gamma-ray interaction (GRI) libraries

  6. Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

    2011-01-01

    Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

  7. Reactor units for power supply to the Russian Arctic regions: Priority assessment of nuclear energy sources

    Directory of Open Access Journals (Sweden)

    Mel'nikov N. N.

    2017-03-01

    Full Text Available Under conditions of competitiveness of small nuclear power plants (SNPP and feasibility of their use to supply power to remote and inaccessible regions the competition occurs between nuclear energy sources, which is caused by a wide range of proposals for solving the problem of power supply to different consumers in the decentralized area of the Russian Arctic power complex. The paper suggests a methodological approach for expert assessment of the priority of small power reactor units based on the application of the point system. The priority types of the reactor units have been determined based on evaluation of the unit's conformity to the following criteria: the level of referentiality and readiness degree of reactor units to implementation; duration of the fuel cycle, which largely determines an autonomy level of the nuclear energy source; the possibility of creating a modular block structure of SNPP; the maximum weight of a transported single equipment for the reactor unit; service life of the main equipment. Within the proposed methodological approach the authors have performed a preliminary ranking of the reactor units according to various criteria, which allows quantitatively determining relative difference and priority of the small nuclear power plants projects aimed at energy supply to the Russian Arctic. To assess the sensitivity of the ranking results to the parameters of the point system the authors have observed the five-point and ten-point scales under variations of importance (weights of different criteria. The paper presents the results of preliminary ranking, which have allowed distinguishing the following types of the reactor units in order of their priority: ABV-6E (ABV-6M, "Uniterm" and SVBR-10 in the energy range up to 20 MW; RITM-200 (RITM-200M, KLT-40S and SVBR-100 in the energy range above 20 MW.

  8. The low power miniature neutron source reactors: Design, safety and applications

    International Nuclear Information System (INIS)

    Ahmed, Y.A.; Ewa, I.O.B.; Umar, M.; Bezboruah, T.; Johri, M.; Akaho, E.H.K.

    2006-04-01

    The Chinese Miniature Neutron Source Reactor (MNSR) is a low power research reactor with maximum thermal neutron flux of 1 x 10 12 n.cm -2 .s -1 in one of its inner irradiation channels and thermal power of approximately 30kW. The MNSR is designed based on the Canadian SLOWPOKE reactor and is one of the smallest commercial research reactors presently available in the world. Its commercial versions currently in operation in China, Ghana, Iran, Nigeria, Pakistan and Syria, is considered as an excellent tool for Neutron Activation Analysis (NAA), training of Scientist, and Engineers in nuclear science and technology and small scale radioisotope production. The paper highlights the basic design and theory of the commercial MNSR, its safety features, applications and advantages over the Chinese Prototype. The experimental flux characteristics determined in this work and in similar studies by other authors reveal that the commercial MNSR has more flux stability, longer life span, higher negative temperature coefficient of reactivity and low under-moderation compared to its prototype in China. The result shows that the facility is safe for reactor physics experiments, teaching and training of students and also ideal for application of NAA for the determination of elemental composition of biological and environmental samples. It can also be a useful tool for geochemical and soil fertility mapping. (author)

  9. Calculations of neutron source at the KYIV research reactor for the boron neutron capture therapy aims

    International Nuclear Information System (INIS)

    Gritzay, O.; Kalchenko, O.; Klimova, N.; Razbudey, V.; Sanzhur, A.

    2006-01-01

    Calculation results of an epithermal neutron source which can be created at the Kyiv Research Reactor (KRR) by means of placing of specially selected moderators, filters, collimators, and shielding into the 10-th horizontal experimental tube (so-called thermal column) are presented. The general Monte-Carlo radiation transport code MCNP4C [1], the Oak Ridge isotope generation code ORIGEN2 [2] and the NJOY99 [3] nuclear data processing system have been used for these calculations

  10. Iser: an international inherently safe reactor concept

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1988-01-01

    Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

  11. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  12. Comparison of SANS instruments at reactors and pulsed sources

    International Nuclear Information System (INIS)

    Thiyagarajan, P.; Epperson, J.E.; Crawford, R.K.; Carpenter, J.M.; Hjelm, R.P. Jr.

    1992-01-01

    Small angle neutron scattering is a general purpose technique to study long range fluctuations and hence has been applied in almost every field of science for material characterization. SANS instruments can be built at steady state reactors and at the pulsed neutron sources where time-of-flight (TOF) techniques are used. The steady state instruments usually give data over small q ranges and in order to cover a large q range these instruments have to be reconfigured several times and SANS measurements have to be made. These instruments have provided better resolution and higher data rates within their restricted q ranges until now, but the TOF instruments are now developing to comparable performance. The TOF-SANS instruments, by using a wide band of wavelengths, can cover a wide dynamic q range in a single measurement. This is a big advantage for studying systems that are changing and those which cannot be exactly reproduced. This paper compares the design concepts and performances of these two types of instruments

  13. Determination of noise sources and space-dependent reactor transfer functions from measured output signals only

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; van Dam, H.; Kleiss, E.B.J.; van Uitert, G.C.; Veldhuis, D.

    1982-01-01

    The measured cross power spectral densities of the signals from three neutron detectors and the displacement of the control rod of the 2 MW research reactor HOR at Delft have been used to determine the space-dependent reactor transfer function, the transfer function of the automatic reactor control system and the noise sources influencing the measured signals. From a block diagram of the reactor with control system and noise sources expressions were derived for the measured cross power spectral densities, which were adjusted to satisfy the requirements following from the adopted model. Then for each frequency point the required transfer functions and noise sources could be derived. The results are in agreement with those of autoregressive modelling of the reactor control feed-back loop. A method has been developed to determine the non-linear characteristics of the automatic reactor control system by analysing the non-gaussian probability density function of the power fluctuations.

  14. Determination of noise sources and space-dependent reactor transfer functions from measured output signals only

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1982-01-01

    The measured cross power spectral densities of the signals from three neutron detectors and the displacement of the control rod of the 2 MW research reactor HOR at Delft have been used to determine the space-dependent reactor transfer function, the transfer function of the automatic reactor control system and the noise sources influencing the measured signals. From a block diagram of the reactor with control system and noise sources expressions were derived for the measured cross power spectral densities, which were adjusted to satisfy the requirements following from the adopted model. Then for each frequency point the required transfer functions and noise sources could be derived. The results are in agreement with those of autoregressive modelling of the reactor control feed-back loop. A method has been developed to determine the non-linear characteristics of the automatic reactor control system by analysing the non-gaussian probability density function of the power fluctuations. (author)

  15. Containment performance analyses for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.

    1992-10-01

    This paper discusses salient aspects of methodology, assumptions, and modeling of various features related to estimation of source terms from two conservatively scoped severe accident scenarios in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for steaming-pool-type accidents and an accident involving molten core-concrete interaction. Several design features (such as rupture disks) are examined to study containment response during postulated severe accidents. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms for each scenario, which are to be used for studying off-site radiological consequences and health effects for these postulated severe accidents. Also highlighted will be a comparison of source terms estimated by two different versions of the MELCOR code

  16. Development of an automatic reactor inspection system

    International Nuclear Information System (INIS)

    Kim, Jae Hee; Eom, Heung Seop; Lee, Jae Cheol; Choi, Yoo Raek; Moon, Soon Seung

    2002-02-01

    Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

  17. Neutron and gamma ray streaming experiments at the fast neutron source reactor 'YAYOI'

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Yanagisawa, Ichiro; Akiyama, Masatsugu; An, Shigehiro

    1979-07-01

    Neutron and gamma ray streaming experiments were performed in the ducts and cavities that were located in the heavy concrete shields of the fast neutron source reactor YAYOI of University of Tokyo. The configurations have the feature that the streaming through the ducts are occurred following the scattering in the cavity. The axes of the ducts are perpendicular to the source radiation from the core. The spectrum of the source was modified by putting a plug in the beam hole of the core. An aluminum plug and the plug which contains paraffin were used. The decay in the ducts, however, hardly depends on the source spectrum. The decay in the ducts is nearly exponential. (author)

  18. Challenges regarding the start-up of an anaerobic biological sulphate reactor using H2 and CO2 as electron donor and carbon sources

    CSIR Research Space (South Africa)

    Roux, SP

    2009-05-01

    Full Text Available Include neutralization and chemical precipitation as well as membrane dependent processes. Biological sulphate reduction is another, environmentally benign option but relies heavily on the availability of an economically viable electron donor...

  19. Analysis of core-concrete interaction event with flooding for the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.

    1993-01-01

    This paper discusses salient aspects of the methodology, assumptions, and modeling of various features related to estimation of source terms from an accident involving a molten core-concrete interaction event (with and without flooding) in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for this postulated severe accident. Several design features (such as rupture disks) are examined to study containment response during this severe accident. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms, which are then used for studying off-site radiological consequences and health effects for the support of the Conceptual Safety Analysis Report for ANS. The results are also to be used to examine the effectiveness of subpile room flooding during this type of severe accident

  20. Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design

    International Nuclear Information System (INIS)

    Wendel, M.W.; Chen, N.C.J.; Yoder, G.L.

    1994-01-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations

  1. Advanced Neutron Source (ANS) Project progress report

    International Nuclear Information System (INIS)

    McBee, M.R.; Chance, C.M.

    1990-04-01

    This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I ampersand C research and development; facility concepts; design; and safety

  2. A conceptual design of neutron tumor therapy reactor facility with a YAYOI based fast neutron source reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki; An, Shigehiro.

    1983-01-01

    Fast neutron is known as one of useful radiations for radiation therapy of tumors. Boron neutron capture therapy (BNCT) of tumors which makes use of 10 B(n, α) 7 Li reaction of 10 B compounds selectively attached to tumor cells with thermal and intermediate neutrons is another way of neutron based radiation therapy which is, above all, attractive enough to kill tumor cells selectively sparing normal tissue. In Japan, BNCT has already been applied and leaned to be effective. After more than a decade operational experiences and the specific experiments designed for therapeutical purposes, in this paper, a conceptual design of a special neutron therapy reactor facility based on YAYOI - fast neutron source reactor of Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo - modified to provide an upward beam of fast and intermediate neutrons is presented. Emphasis is placed on the in-house nature of facility and on the coordinating capability of biological and physical researches as well as maintenances of the facility. (author)

  3. Structure shielding from cloud and fallout gamma ray sources for assessing the consequences of reactor accidents

    International Nuclear Information System (INIS)

    Burson, Z.G.; Profio, A.E.

    1975-12-01

    Radiation shielding provided by transportation vehicles and structures typical of where people live and work were estimated for cloud and fallout gamma-ray sources resulting from a hypothetical reactor accident. Dose reduction factors are recommended for a variety of situations for realistically assessing the consequences of reactor accidents

  4. SEBREZ: an inertial-fusion-reactor concept

    International Nuclear Information System (INIS)

    Meier, W.R.

    1982-01-01

    The neutronic aspects of an inertial fusion reactor concept that relies on asymmetrical neutronic effects to enhance the tritium production in the breeding zones have been studied. We find that it is possible to obtain a tritium breeding ratio greater than 1.0 with a chamber configuration in which the breeding zones subtend only a fraction of the total solid angle. This is the origin of the name SEBREZ which stands for SEgregated BREeding Zones. It should be emphasized that this is not a reactor design study; rather this study illustrates certain neutronic effects in the context of a particular reactor concept. An understanding of these effects forms the basis of a design technique which has broader application than just the SEBREZ concept

  5. Radiation protection programme for LEU miniature source reactor

    International Nuclear Information System (INIS)

    Beinpuo, Ernest Sanyare Warmann

    2015-02-01

    A radiation protection program has been developed to promote radiation dose reduction. It emphasize radiological protection fundamentals geared at reducing radiation from the application of the research reactor at the reactor center of the National Nuclear Research Institute (NNRI) of the Ghana Atomic Energy Commission. The objectives of the radiation safety program are both to ensure that nuclear scientists and technicians are exposed to a minimum of ionizing radiation and to protect employees and facility users and surrounding community from any potentially harmful effects of nuclear research reactor at GAEC. The primary purpose of the radiation control program is to assure radiological safety of all personnel and the public to guarantee that ionizing radiation arising out of the operations of the Research Reactor at the Reactor Center does not adversely affect personnel, the general public or the environment. This program sets forth polices, regulations, and procedures approved by the Centers Radiation Control Committee. The regulations and procedures outlined in this program are intended to protect all individuals with a minimum of interference in their activities and are consistent with regulations of the Radiation Protection Board (RPB) applicable to ionizing radioactive producing devices. (au)

  6. Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

    International Nuclear Information System (INIS)

    Muswema, J.L.; Ekoko, G.B.; Lukanda, V.M.; Lobo, J.K.-K.; Darko, E.O.; Boafo, E.K.

    2015-01-01

    Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

  7. Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

    Energy Technology Data Exchange (ETDEWEB)

    Muswema, J.L., E-mail: jeremie.muswem@unikin.ac.cd [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Ekoko, G.B. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Lukanda, V.M. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Democratic Republic of the Congo' s General Atomic Energy Commission, P.O. Box AE1 (Congo, The Democratic Republic of the); Lobo, J.K.-K. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Darko, E.O. [Radiation Protection Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Boafo, E.K. [University of Ontario Institute of Technology, 2000 Simcoe St. North, Oshawa, ONL1 H7K4 (Canada)

    2015-01-15

    Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

  8. Utilization of vinasse effluents from an anaerobic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Costa, F J.C.B.; Rocha, B B.M.; Viana, C E; Toledo, A C

    1986-01-01

    An anaerobic reactor was developed to biodigest alcohol distillery wastes. A further post-treatment of the effluent reduced the level of pollution to the point of eventually discharging into streams and rivers. The present work also analyses the use of biodigested vinasse as a source of food for fish. Very high efficiencies were obtained during primary and secondary treatment of vinasse effluent, as demonstrated by the greatly reduced organic load. The utilization of the treated effluent as a source of fish food presents an excellent alternative for the Brazilian alcohol industry. (Refs. 6).

  9. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    International Nuclear Information System (INIS)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2016-01-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  10. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  11. The Pebble Bed Modular Reactor: An obituary

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Steve, E-mail: stephen.thomas@gre.ac.u [Public Services International Research Unit (PSIRU), Business School, University of Greenwich, 30 Park Row, London SE10 9LS (United Kingdom)

    2011-05-15

    The High Temperature Gas-cooled Reactor (HTGR) has exerted a peculiar attraction over nuclear engineers. Despite many unsuccessful attempts over half a century to develop it as a commercial power reactor, there is still a strong belief amongst many nuclear advocates that a highly successful HTGR technology will emerge. The most recent attempt to commercialize an HTGR design, the Pebble Bed Modular Reactor (PBMR), was abandoned in 2010 after 12 years of effort and the expenditure of a large amount of South African public money. This article reviews this latest attempt to commercialize an HTGR design and attempts to identify which issues have led to its failure and what lessons can be learnt from this experience. It concludes that any further attempts to develop HTGRs using Pebble Bed technology should only be undertaken if there is a clear understanding of why earlier attempts have failed and a high level of confidence that earlier problems have been overcome. It argues that the PBMR project has exposed serious weaknesses in accountability mechanisms for the expenditure of South African public money. - Research highlights: {yields} In this study we examine the reasons behind the failure of the South African PBMR programme. {yields} The study reviews the technical issues that have arisen and lessons for future reactor developments. {yields} The study also identifies weaknesses in the accountability mechanisms for public spending.

  12. An autonomous control framework for advanced reactors

    Directory of Open Access Journals (Sweden)

    Richard T. Wood

    2017-08-01

    Full Text Available Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

  13. An autonomous control framework for advanced reactors

    International Nuclear Information System (INIS)

    Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C.

    2017-01-01

    Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors

  14. An autonomous control framework for advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

  15. Jet flow analysis of liquid poison injection in a CANDU reactor using source term

    International Nuclear Information System (INIS)

    Chae, Kyung Myung; Choi, Hang Bok; Rhee, Bo Wook

    2001-01-01

    For the performance analysis of Canadian deuterium uranium (CANDU) reactor shutdown system number 2 (SDS2), a computational fluid dynamics model of poison jet flow has been developed to estimate the flow field and poison concentration formed inside the CANDU reactor calandria. As the ratio of calandria shell radius over injection nozzle hole diameter is so large (1055), it is impractical to develop a full-size model encompassing the whole calandria shell. In order to reduce the model to a manageable size, a quarter of one-pitch length segment of the shell was modeled using symmetric nature of the jet; and the injected jet was treated as a source term to avoid the modeling difficulty caused by the big difference of the hole sizes. For the analysis of an actual CANDU-6 SDS2 poison injection, the grid structure was determined based on the results of two-dimensional real- and source-jet simulations. The maximum injection velocity of the liquid poison is 27.8 m/s and the mass fraction of the poison is 8000 ppm (mg/kg). The simulation results have shown well-established jet flow field. In general, the jet develops narrowly at first but stretches rapidly. Then, the flow recirculates a little in r-x plane, while it recirculates largely in r-θ plane. As the time goes on, the adjacent jets contact each other and form a wavy front such that the whole jet develops in a plate form. his study has shown that the source term model can be effectively used for the analysis of the poison injection and the simulation result of the CANDU reactor is consistent with the model currently being used for the safety analysis. In the future, it is strongly recommended to analyze the transient (from helium tank to injection nozzle hole) of the poison injection by applying Bernoulli equation with real boundary conditions

  16. Comparison of xenon-135 and samarium-149 poisoning in the Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-05-01

    The Xe-135 equilibrium reactivity was calculated previously using the WIMSD4 and CITATION codes to estimate the fission product poisoning factor in the Syrian Miniature Neutron Source Reactor (MNSR). The fission product poisoning factor was used to calculate the Xe-135 concentrations and reactivities during the reactor operation, at saturation, and after shutdown. The fission product poisoning factor is used in this study to calculate the Sm-149 concentrations and reactivities during the reactor operation, at saturation, and after shutdown. The results are compared with Xe-135 concentrations and reactivities which were calculated before. The Xe-135 concentration increases versus the reactor operation time and reaches 2.836*10 13 atoms/cm 3 after 6 hours, where the Sm-149 concentration reaches 0.745*10 12 atoms/cm 3 . The Xe-135 and Sm-149 reactivities reach -0.451 and -0.256*10 -3 mk respectively after 6 hours of the reactor operation time. It is found that the Xe-135 and Sm-149 reactivities do not reach the equilibrium reactivities during typical reactor daily operating time. The Xe-235 concentration and reactivity are increased and reach to maximum values after 30000 second from the reactor shut down. Its maximum reactivity is -0.903 mk. Hence, the Xe-235 generates -0.452 mk extra reactivity after the reactor shutdown. The Sm-149 concentration and reactivates are increased after the reactor shut down. The Sm-149 reactivity reaches -0.249*10 -2 mk after 100000 second from the reactor shut down, and the Sm-149 introduces -2.231*10 -3 mk extra reactivity after the reactor shut down. Finally, the amount of Sm-149 accumulated in the reactor core after 1, 2, 5, and 10 years of the reactor operation time are: 3.995*10 -3 , 7.694*10 -3 , 17.226*10 -3 , and 2.892*10 -2 grams respectively. (Author)

  17. The safety features of an integrated maritime reactor

    International Nuclear Information System (INIS)

    Miyakoshi, Junichi; Yamada, Nobuyuki; Kuwahara, Shin-ichi

    1975-01-01

    The EFDR-80, a typical integrated maritime reactor, which is being developed in West Germany is outlined. The safety features of the integrated maritime reactor are presented with the analysis of reactor accidents and hazards, and are compared with those of the separated maritime reactor. Furthermore, the safety criteria of maritime reactors in Japan and West Germany are compared, and some of the differences are presented from the viewpoint of reactor design and safety analysis. In this report the authors express an earnest desire that the definite and reasonable safety criteria of the integrated maritime reactor should be established and that the safety criteria of the nuclear ship should be standardized internationally. (auth.)

  18. Effect of hypoiodous acid volatility on the iodine source term in reactor accidents

    International Nuclear Information System (INIS)

    Routamo, T.

    1996-01-01

    A FORTRAN code ACT WATCH has been developed to establish an improved understanding of essential radionuclide behaviour mechanisms, especially related to iodine chemistry, in reactor accidents. The accident scenarios calculated in this paper are based on the Loss of Coolant accident at the Loviisa Nuclear Power Plant. The effect of different airborne species, especially HIO, on the iodine source term has been studied. The main cause of the high HIO release in the system modelled is the increase of I 2 hydrolysis rate along with the temperature increase, which accelerates HIO production. Due to the high radiation level near the reactor core, I 2 is produced from I - very rapidly. High temperature in the reactor coolant causes I 2 to be transformed into HIO and through the boiling of the coolant volatile I 2 and HIO are transferred efficiently into the gas phase. High filtration efficiency for particulate iodine causes I - release to be much lower than those of I 2 and HIO. (author) 15 figs., 1 tab., refs

  19. Effect of hypoiodous acid volatility on the iodine source term in reactor accidents

    Energy Technology Data Exchange (ETDEWEB)

    Routamo, T [Imatran Voima Oy, Vantaa (Finland)

    1996-12-01

    A FORTRAN code ACT WATCH has been developed to establish an improved understanding of essential radionuclide behaviour mechanisms, especially related to iodine chemistry, in reactor accidents. The accident scenarios calculated in this paper are based on the Loss of Coolant accident at the Loviisa Nuclear Power Plant. The effect of different airborne species, especially HIO, on the iodine source term has been studied. The main cause of the high HIO release in the system modelled is the increase of I{sub 2} hydrolysis rate along with the temperature increase, which accelerates HIO production. Due to the high radiation level near the reactor core, I{sub 2} is produced from I{sup -}very rapidly. High temperature in the reactor coolant causes I{sub 2} to be transformed into HIO and through the boiling of the coolant volatile I{sub 2} and HIO are transferred efficiently into the gas phase. High filtration efficiency for particulate iodine causes I{sup -} release to be much lower than those of I{sub 2} and HIO. (author) 15 figs., 1 tab., refs.

  20. Computer modelling of radioactive source terms at a tokamak reactor

    International Nuclear Information System (INIS)

    Meide, A.

    1984-12-01

    The Monte Carlo code MCNP has been used to create a simple three-dimensional mathematical model representing 1/12 of a tokamak fusion reactor for studies of the exposure rate level from neutrons as well as gamma rays from the activated materials, and for later estimates of the consequences to the environment, public, and operating personnel. The model is based on the recommendations from the NET/INTOR workshops. (author)

  1. The SILENE reactor: an instrument suitable for studying the impact of intermediate and high radiation doses

    International Nuclear Information System (INIS)

    Verray, B.; Leo, Y.; Fouillaud, P.

    2002-01-01

    Designed in 1974 to study the phenomenology and consequences of a criticality accident, the SILENE experimental reactor, an intense source of mixed neutron and gamma radiation, is also suited to radiobiological studies. (author)

  2. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, Murray Wilford [ORNL

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  3. The High Flux Isotope Reactor (HFIR) cold source project at ORNL

    International Nuclear Information System (INIS)

    Selby, D.L.; Lucas, A.T.; Chang, S.J.; Freels, J.D. . E-mail-yb2@ornl.gov

    1998-01-01

    Following the decision to cancel the Advanced Neutron Source (ANS) Project at Oak Ridge National Laboratory (ORNL), it was determined that a hydrogen cold source should be retrofitted into an existing beam tube of the High Flux Isotope Reactor (HFIR) at ORNL. The preliminary design of this system has been completed and an 'approval in principle' of the design has been obtained from the internal ORNL safety review committees and the U.S. Department of Energy (DOE) safety review committee. The cold source concept is basically a closed loop forced flow supercritical hydrogen system. The supercritical approach was chosen because of its enhanced stability in the proposed high heat flux regions. Neutron and gamma physics of the moderator have been analyzed using the 3D Monte Carlo code MCNP 1 A D structural analysis model of the moderator vessel, vacuum tube, and beam tube was completed to evaluate stress loadings and to examine the impact of hydrogen detonations in the beam tube. A detailed ATHENA 2 system model of the hydrogen system has been developed to simulate loop performance under normal and off-normal transient conditions. Semi-prototypic hydrogen loop tests of the system have been performed at the Arnold Engineering Design Center (AEDC) located in Tullahoma, Tennessee to verify the design and benchmark the analytical system model. A 3.5 kW refrigerator system has been ordered and is expected to be delivered to ORNL by the end of this calendar year. Our present schedule shows the assembling of the cold source loop on site during the fall of 1999 for final testing before insertion of the moderator plug assembly into the reactor beam tube during the end of the year 2000. (author)

  4. Development of source term evaluation method for Korean Next Generation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Keon Jae; Cheong, Jae Hak; Park, Jin Baek; Kim, Guk Gee [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-10-15

    This project had investigate several design features of radioactive waste processing system and method to predict nuclide concentration at primary coolant basic concept of next generation reactor and safety goals at the former phase. In this project several prediction methods of source term are evaluated conglomerately and detailed contents of this project are : model evaluation of nuclide concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant(NPP), investigation of prediction parameter of source term evaluation, basic parameter of PWR, operational parameter, respectively, radionuclide removal system and adjustment values of reference NPP, suggestion of source term prediction method of next generation NPP.

  5. Production of the sealed gamma-radiation sources of with iridium-192 radionuclide at the WWR-K research reactor

    International Nuclear Information System (INIS)

    Petukhov, V.K.; Chernayev, V.P.; Chabeyev, N.T.; Ermakov, E.L.; Chakrov, P.V.

    2005-01-01

    Full text: Conversion orientation of the WWR-K research reactor activity was established after renewal of its operation in 1997. A priority in reactor works was determined in the decision of tasks of practical use of nuclear technologies in a national economy in the next directions: in an industry, public health services and agriculture. The items of prime tasks: development and introduction of radiation technologies and manufacturing of radioisotopes for industry. This task included both scientific and technical program in the list of works of the Republican goals. At the WWR-K reactor within the framework of the this task solution the works on pilot production of the sealed sources of radioactive radiations (SSRR) with Ir-192 radionuclide for an industry of Republic of Kazakhstan were made. Organizational questions related to the Kazakhstan authority body and the regulating documentation were solved the first of all. The second stage was the development of the techniques of creating of devices providing an samples irradiation in reactor, control of sources sealing, measurements of the equivalent radiation doze from sources and high-quality support of SSRR manufacture over all technological way. At the third stage was made a little quantity SSRR with Ir-192 radionuclide, such as GIID-A1 (G6), for 'TEKOPS-660' Gammaray Projectors. This work served as experimental check of the decisions correctness, and has allowed to remove those lacks, to find out which it was possible only during direct manufacturing of radioactive sources. During performance of all these works the following was carried out: development and release of the documents and specifications regulating work on SSRR manufacture at the Institute of Nuclear Physics; personnel preparation and certification; preparation and equipment providing of reactor hot chambers by additional devices for work with irradiated iridium samples; development and manufacturing of the devices for iridium samples irradiation in

  6. The design of the cold neutron source of the OPAL reactor

    International Nuclear Information System (INIS)

    Rechiman, L.M.; Bonetto, Fabian J.; Buscaglia, Gustavo C.

    2007-01-01

    The present work describes the conceptual design process of the first cold neutron source developed by INVAP for the nuclear research reactor OPAL. The analysis begins from the requirements given by the client and continues with the chosen solutions. Furthermore, we studied how impact in the design the fully illuminated constraint with the finite remote source model. (author) [es

  7. Application of SAFE to an operating reactor

    International Nuclear Information System (INIS)

    Chapman, L.D.

    1979-01-01

    A method for the evaluation of physical protection systems at nuclear facilities has been developed. The evaluation process consists of five major phases: (1) Facility Characterization, (2) Facility Representation, (3) Component Performance, (4) Adversary Path Analysis, and (5) Effectiveness Evaluation. Each of these phases will be described in some detail and illustrated by examples. The process for evaluation of physical protection system effectiveness against an outside threat will be presented for a reactor facility

  8. An Accelerator Neutron Source for BNCT

    International Nuclear Information System (INIS)

    Blue, Thomas E.

    2006-01-01

    The overall goal of this project was to develop an accelerator-based neutron source (ABNS) for Boron Neutron Capture Therapy (BNCT). Specifically, our goals were to design, and confirm by measurement, a target assembly and a moderator assembly that would fulfill the design requirements of the ABNS. These design requirements were (1) that the neutron field quality be as good as the neutron field quality for the reactor-based neutron sources for BNCT, (2) that the patient treatment time be reasonable, (3) that the proton current required to treat patients in reasonable times be technologically achievable at reasonable cost with good reliability, and accelerator space requirements which can be met in a hospital, and finally (4) that the treatment be safe for the patients

  9. An Accelerator Neutron Source for BNCT

    Energy Technology Data Exchange (ETDEWEB)

    Blue, Thomas, E

    2006-03-14

    The overall goal of this project was to develop an accelerator-based neutron source (ABNS) for Boron Neutron Capture Therapy (BNCT). Specifically, our goals were to design, and confirm by measurement, a target assembly and a moderator assembly that would fulfill the design requirements of the ABNS. These design requirements were 1) that the neutron field quality be as good as the neutron field quality for the reactor-based neutron sources for BNCT, 2) that the patient treatment time be reasonable, 3) that the proton current required to treat patients in reasonable times be technologially achievable at reasonable cost with good reliability, and accelerator space requirements which can be met in a hospital, and finally 4) that the treatment be safe for the patients.

  10. Ion source development for a photoneutralization based NBI system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simonin, A.; Esch, H. P. L. de; Garibaldi, P.; Grand, C.; Bechu, S.; Bès, A.; Lacoste, A. [CEA-Cadarache, IRFM, F-13108 St. Paul-lez-Durance (France); LPSC, Grenoble-Alpes University, F-38026 Grenoble France (France)

    2015-04-08

    The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D° beam at 1MeV, including high wall-plug efficiency (η > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D{sup −} beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R and D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.

  11. Ion source development for a photoneutralization based NBI system for fusion reactors

    International Nuclear Information System (INIS)

    Simonin, A.; Esch, H. P. L. de; Garibaldi, P.; Grand, C.; Bechu, S.; Bès, A.; Lacoste, A.

    2015-01-01

    The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D° beam at 1MeV, including high wall-plug efficiency (η > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D − beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R and D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities

  12. Advanced Neutron Source Reactor zoning, shielding, and radiological optimization guide

    International Nuclear Information System (INIS)

    Westbrook, J.L.; DeVore, J.R.

    1995-08-01

    In the design of major nuclear facilities, it is important to protect both humans and equipment excessive radiation dose. Past experience has shown that it is very effective to apply dose reduction principles early in the design of a nuclear facility both to specific design features and to the manner of operation of the facility, where they can aid in making the facility more efficient and cost-effective. Since the appropriate choice of radiological controls and practices varies according to the case, each area of the facility must be analyzed for its radiological impact, both by itself and in interactions with other areas. For the Advanced Neutron Source (ANS) project, a large relational database will be used to collect facility information by system and relate it to areas. The database will also hold the facility dose and shielding information as it is produced during the design process. This report details how the ANS zoning scheme was established and how the calculation of doses and shielding are to be done

  13. The effect of toxic carbon source on the reaction of activated sludge in the batch reactor.

    Science.gov (United States)

    Wu, Changyong; Zhou, Yuexi; Zhang, Siyu; Xu, Min; Song, Jiamei

    2018-03-01

    The toxic carbon source can cause higher residual effluent dissolved organic carbon than easily biodegraded carbon source in activated sludge process. In this study, an integrated activated sludge model is developed as the tool to understand the mechanism of toxic carbon source (phenol) on the reaction, regarding the carbon flows during the aeration period in the batch reactor. To estimate the toxic function of phenol, the microbial cells death rate (k death ) is introduced into the model. The integrated model was calibrated and validated by the experimental data and it was found the model simulations matched the all experimental measurements. In the steady state, the toxicity of phenol can result in higher microbial cells death rate (0.1637 h -1 vs 0.0028 h -1 ) and decay rate coefficient of biomass (0.0115 h -1 vs 0.0107 h -1 ) than acetate. In addition, the utilization-associated products (UAP) and extracellular polymeric substances (EPS) formation coefficients of phenol are higher than that of acetate, indicating that more carbon flows into the extracellular components, such as soluble microbial products (SMP), when degrading toxic organics. In the non-steady state of feeding phenol, the yield coefficient for growth and maximum specific growth rate are very low in the first few days (1-10 d), while the decay rate coefficient of biomass and microbial cells death rate are relatively high. The model provides insights into the difference of the dynamic reaction with different carbon sources in the batch reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1982-01-01

    REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

  15. Condensed matter and materials research using neutron diffraction and spectroscopy: reactor and pulsed neutron sources

    International Nuclear Information System (INIS)

    Bisanti, Paola; Lovesey, S.W.

    1987-05-01

    The paper provides a short, and partial view of the neutron scattering technique applied to condensed matter and materials research. Reactor and accelerator-based neutron spectrometers are discussed, together with examples of research projects that illustrate the puissance and modern applications of neutron scattering. Some examples are chosen to show the range of facilities available at the medium flux reactor operated by Casaccia ENEA, Roma and the advanced, pulsed spallation neutron source at the Rutherford Appleton Laboratory, Oxfordshire. (author)

  16. The source term and waste optimization of molten salt reactors with processing

    International Nuclear Information System (INIS)

    Gat, U.; Dodds, H.L.

    1993-01-01

    The source term of a molten salt reactor (MSR) with fuel processing is reduced by the ratio of processing time to refueling time as compared to solid fuel reactors. The reduction, which can be one to two orders of magnitude, is due to removal of the long-lived fission products. The waste from MSRs can be optimized with respect to its chemical composition, concentration, mixture, shape, and size. The actinides and long-lived isotopes can be separated out and returned to the reactor for transmutation. These features make MSRs more acceptable and simpler in operation and handling

  17. Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

  18. Derivation of the source term, dose results and associated radiological consequences for the Greek Research Reactor – 1

    Energy Technology Data Exchange (ETDEWEB)

    Pappas, Charalampos, E-mail: chpappas@ipta.demokritos.gr; Ikonomopoulos, Andreas; Sfetsos, Athanasios; Andronopoulos, Spyros; Varvayanni, Melpomeni; Catsaros, Nicolas

    2014-07-01

    Highlights: • Source term derivation of postulated accident sequences in a research reactor. • Various containment ventilation scenarios considered for source term calculations. • Source term parametric analysis performed in case of lack of ventilation. • JRODOS employed for dose calculations under eighteen modeled scenarios. • Estimation of radiological consequences during typical and adverse weather scenarios. - Abstract: The estimated source term, dose results and radiological consequences of selected accident sequences in the Greek Research Reactor – 1 are presented and discussed. A systematic approach has been adopted to perform the necessary calculations in accordance with the latest computational developments and IAEA recommendations. Loss-of-coolant, reactivity insertion and fuel channel blockage accident sequences have been selected to derive the associated source terms under three distinct containment ventilation scenarios. Core damage has been conservatively assessed for each accident sequence while the ventilation has been assumed to function within the efficiency limits defined at the Safety Analysis Report. In case of lack of ventilation a parametric analysis is also performed to examine the dependency of the source term on the containment leakage rate. A typical as well as an adverse meteorological scenario have been defined in the JRODOS computational platform in order to predict the effective, lung and thyroid doses within a region defined by a 15 km radius downwind from the reactor building. The radiological consequences of the eighteen scenarios associated with the accident sequences are presented and discussed.

  19. Consequences of the conversion of research reactor cores on experimental facilities at the example of a cold neutron source

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Goebs, H.; Stroemich, A.

    1985-01-01

    The consequences for and specifically the potential reduction of the performance of research reactors have been in discussions very often within the last five years as one of the draw-backs which has to be paid for the reduction of the proliferation risk at research reactor plants. Up to now and up to our knowledge the available results are restricted to unperturbated fluxes. Thus, this contribution makes the attempt to demonstrate the consequence of core conversion on an example of a real experimental facility and - at the same time - on one that is going to be used in the next decade a lot, i.e. a cold neutron source (CNS). (author)

  20. Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.

    2017-06-26

    Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

  1. Performance of an innovative multi-stage anaerobic reactor during ...

    African Journals Online (AJOL)

    Start-up of an anaerobic reactor is a relatively delicate process and depends on various factors such as wastewater composition, available inoculum, operating conditions and reactor configuration. Accordingly, systematized operational procedures are important, mainly during the start-up of an anaerobic reactor.

  2. A new oxidation flow reactor for measuring secondary aerosol formation of rapidly changing emission sources

    Science.gov (United States)

    Simonen, Pauli; Saukko, Erkka; Karjalainen, Panu; Timonen, Hilkka; Bloss, Matthew; Aakko-Saksa, Päivi; Rönkkö, Topi; Keskinen, Jorma; Dal Maso, Miikka

    2017-04-01

    Oxidation flow reactors (OFRs) or environmental chambers can be used to estimate secondary aerosol formation potential of different emission sources. Emissions from anthropogenic sources, such as vehicles, often vary on short timescales. For example, to identify the vehicle driving conditions that lead to high potential secondary aerosol emissions, rapid oxidation of exhaust is needed. However, the residence times in environmental chambers and in most oxidation flow reactors are too long to study these transient effects ( ˜ 100 s in flow reactors and several hours in environmental chambers). Here, we present a new oxidation flow reactor, TSAR (TUT Secondary Aerosol Reactor), which has a short residence time ( ˜ 40 s) and near-laminar flow conditions. These improvements are achieved by reducing the reactor radius and volume. This allows studying, for example, the effect of vehicle driving conditions on the secondary aerosol formation potential of the exhaust. We show that the flow pattern in TSAR is nearly laminar and particle losses are negligible. The secondary organic aerosol (SOA) produced in TSAR has a similar mass spectrum to the SOA produced in the state-of-the-art reactor, PAM (potential aerosol mass). Both reactors produce the same amount of mass, but TSAR has a higher time resolution. We also show that TSAR is capable of measuring the secondary aerosol formation potential of a vehicle during a transient driving cycle and that the fast response of TSAR reveals how different driving conditions affect the amount of formed secondary aerosol. Thus, TSAR can be used to study rapidly changing emission sources, especially the vehicular emissions during transient driving.

  3. Identification of neutron noise sources in a boiling water reactor

    International Nuclear Information System (INIS)

    Sides, W.H. Jr.; Mathis, M.V.; Smith, C.M.

    1977-01-01

    Measurements were made at units 2 and 3 of the Browns Ferry Nuclear Power Plant in order to characterize the noise signatures of the neutron and process signals and to determine the usefulness of such signatures for anomaly detection in BWR-4s. Previous measurements and theoretical analyses of BWR noise by others were concerned with the determination of steam velocity and void fraction (using the local component of neutron noise) and with the sources of global noise. The work described is under a five-part program to develop a complete and systematic analysis and representation of BWR neutron and process noise through complementary measurements and stochastic model developments. The parts are: (1) recording as many neutron detector and process noise signals as are available in a BWR-4; (2) reducing these data to noise signatures in order to perform an empirical analysis of these signatures, and documenting the relationships between the signals from spatially separated neutron detectors and between neutron and process variables; (3) developing spatially dependent neutronic models coupled with thermal-hydraulic models to aid in interpreting the observed relationships among the measured noise signatures, (4) comparing measured noise signatures with model predictions to obtain additional insight into BWR-4 dynamic behavior and to validate the models; and (5) using these models to predict the sensitivity of noise monitoring for detection, surveillance, and diagnosis of postulated in-core anomalies in BWRs. The paper describes the procedures used to obtain the noise recordings and presents initial empirical analysis and observations pertaining to the noise signatures and the relationships between several noise variables in the 0.01- to 1-Hz range. The mathematical models have not been developed sufficiently to report theoretical results or to compare measured spectra with model predictions at this time

  4. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    Schulz, K.C.

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K Q due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

  5. Representativeness elements of an hybrid reactor demonstrator

    International Nuclear Information System (INIS)

    Kerdraon, D.; Billebaud, A.; Brissot, R.; David, S.; Giorni, A.; Heuer, D.; Loiseaux, J.M.; Meplan, O.

    2000-11-01

    This document deals with the quantification of the minimum thermal power level for a demonstrator and the definition of the physical criteria which define the representative character of a demonstrator towards a power reactor. Solutions allowing to keep an acceptable flow in an industrial core, have also been studied. The document is divided in three parts: the representativeness elements, the considered solutions and the characterization of the neutrons flows at the interfaces and the dose rates at the outer surface of the vessel. (A.L.B.)

  6. TRIGA reactor as an experimental tool

    Energy Technology Data Exchange (ETDEWEB)

    Nahrul Khair bin Alang Mohammad Rashid (PUSPATI, Selangor (Malaysia))

    1981-01-01

    Article reviewed on the general features, operation and capabilities, and utilization of a research reactor, PUSPATI TRIGA MARK II. The paper also described the arrangements for the use of the PUSPATI reactor.

  7. Triga reactor as an experimental tool

    International Nuclear Information System (INIS)

    Nahrul Khair bin Alang Mohammad Rashid

    1981-01-01

    Article reviewed on the general features, operation and capabilities, and utilization of a research reactor, PUSPATI TRIGA MARK II. The paper also described the arrangements for the use of the PUSPATI reactor

  8. Enhanced biological phosphorus removal in a sequencing batch reactor using propionate as the sole carbon source.

    Science.gov (United States)

    Pijuan, M; Saunders, A M; Guisasola, A; Baeza, J A; Casas, C; Blackall, L L

    2004-01-05

    An enhanced biological phosphorus removal (EBPR) system was developed in a sequencing batch reactor (SBR) using propionate as the sole carbon source. The microbial community was followed using fluorescence in situ hybridization (FISH) techniques and Candidatus 'Accumulibacter phosphatis' were quantified from the start up of the reactor until steady state. A series of SBR cycle studies was performed when 55% of the SBR biomass was Accumulibacter, a confirmed polyphosphate accumulating organism (PAO) and when Candidatus 'Competibacter phosphatis', a confirmed glycogen-accumulating organism (GAO), was essentially undetectable. These experiments evaluated two different carbon sources (propionate and acetate), and in every case, two different P-release rates were detected. The highest rate took place while there was volatile fatty acid (VFA) in the mixed liquor, and after the VFA was depleted a second P-release rate was observed. This second rate was very similar to the one detected in experiments performed without added VFA.A kinetic and stoichiometric model developed as a modification of Activated Sludge Model 2 (ASM2) including glycogen economy, was fitted to the experimental profiles. The validation and calibration of this model was carried out with the cycle study experiments performed using both VFAs. The effect of pH from 6.5 to 8.0 on anaerobic P-release and VFA-uptake and aerobic P-uptake was also studied using propionate. The optimal overall working pH was around 7.5. This is the first study of the microbial community involved in EBPR developed with propionate as a sole carbon source along with detailed process performance investigations of the propionate-utilizing PAOs. Copyright 2003 Wiley Periodicals, Inc.

  9. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided

  10. Using level-I PRA for enhanced safety of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    Ramsey, C.T.; Linn, M.A.

    1995-01-01

    The phase-1, level-I probabilistic risk assessment (PRA) of the Advanced Neutron Source (ANS) reactor has been completed as part of the conceptual design phase of this proposed research facility. Since project inception, PRA and reliability concepts have been an integral part of the design evolutions contributing to many of the safety features in the current design. The level-I PRA has been used to evaluate the internal events core damage frequency against project goals and to identify systems important to safety and availability, and it will continue to guide and provide support to accident analysis, both severe and nonsevere. The results also reflect the risk value of defense-in-depth safety features in reducing the likelihood of core damage

  11. Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-01-01

    A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios

  12. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  13. Fusion: an energy source for synthetic fuels

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J; Steinberg, M.

    1980-01-01

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and supplement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of approx. 40 to 60% and hydrogen production efficiencies by high temperature electrolysis of approx. 50 to 70% are projected for fusion reactors using high temperature blankets. Fusion/coal symbiotic systems appear economically promising for the first generation of commercial fusion synfuels plants. Coal production requirements and the environmental effects of large-scale coal usage would be greatly reduced by a fusion/coal system. In the long term, there could be a gradual transition to an inexhaustible energy system based solely on fusion

  14. A comparison of world-wide uses of severe reactor accident source terms

    International Nuclear Information System (INIS)

    Ang, M.L.; Frid, W.; Kersting, E.J.; Friederichs, H.G.; Lee, R.Y.; Meyer-Heine, A.; Powers, D.A.; Soda, K.; Sweet, D.

    1994-09-01

    The definitions of source terms to reactor containments and source terms to the environment are discussed. A comparison is made between the TID-14844 example source term and the alternative source term described in NUREG-1465. Comparisons of these source terms to the containments and those used in France, Germany, Japan, Sweden, and the United Kingdom are made. Source terms to the environment calculated in NUREG-1500 and WASH-1400 are discussed. Again, these source terms are compared to those now being used in France, Germany, Japan, Sweden, and the United Kingdom. It is concluded that source terms to the containment suggested in NUREG-1465 are not greatly more conservative than those used in other countries. Technical bases for the source terms are similar. The regulatory use of the current understanding of radionuclide behavior varies among countries

  15. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

    2011-03-01

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one

  16. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Bryant, Rebecca; Kszos, Lynn A.

    2011-01-01

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews

  17. An energy amplifier fluidized bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, F.; Seifritz, W.

    2001-01-01

    The concept of a fluidized bed nuclear reactor driven by an energy amplifier system is described. The reactor has promising characteristics of inherent safety and passive cooling. The reactor can easily operate with any desired spectrum in order to be a plutonium burner or have it operate with thorium fuel cycle. (orig.) [de

  18. Researches at the University of Tokyo fast neutron sources reactor, YAYOI

    International Nuclear Information System (INIS)

    Koshizuka, S.; Oka, Y.; Saito, I.

    1992-01-01

    The Fast neutron source reactor YAYOI was critical in 1971 at the Nuclear Engineering Research Laboratory, the Faculty of Engineering, the University of Tokyo (UTNL). The core is fueled with the enriched uranium surrounded by the depleted uranium. YAYOI is the first fast reactor in Japan. Many types of studies have been carried out by the researchers of the University of Tokyo in these 20 years. It also contributed to the Japan's national project of developing fast breeder reactors. The reactor is opened to the visiting researchers from universities and research institutes. YAYOI has also been utilized for education of undergraduate and graduate students of the Department of Nuclear Engineering of the University of Tokyo. The present paper briefly summerizes past and present researchers. (author)

  19. An appropriate level for reactor regulatory research

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Kouts, H.J.C.

    1986-01-01

    In this brief paper the appropriate role for NRC's research program is set down, based on the experience of the last decade; and the broad outlines of an appropriate research program for the next decade are discussed. The two authors bring to this topic direct personal experience: they are two of only three former directors of NRC's research program in its history (the other was the last Saul Levine). The discussion considers only reactor safety research, although NRC's research program covers several other areas as well. Safeguards, environmental and health studies, waste management, and transportation are not covered

  20. An introduction to nuclear reactor theory

    International Nuclear Information System (INIS)

    Iliffe, C.E.

    1984-01-01

    The book is intended primarily for those who go into the design, development and operation of nuclear power plants from other industries. Thus no prior knowledge of nuclear power or its underlying physical principles is assumed. The aim is to complement existing text books on reactor theory. The 23 chapters start with the earliest concepts of atomic structure and trace an historical sequence of theoretical and practical achievement to the realisation of nuclear power. In doing this some original references not previously published in books are used. (U.K.)

  1. RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

  2. A new method to determine in situ the transmission of a neutron-guide system at a reactor source

    CERN Document Server

    Haan, V O D; Gommers, R M; Labohm, F; Well, A A V; De Leege, P F A; Schebetov, A; Pusenkov, V

    2002-01-01

    In this paper, a description of a new method to determine the transmission of neutron guides after they are installed in a beam-tube at a reactor source is given. The method is based on activation measurements of gold foils at the entrance of the beam-tube and at the exit of the neutron guides compared to Monte-Carlo calculations. In this method, a quality factor is defined as the ratio between the actual transmission and the theoretical maximum attainable transmission. This method is used to determine the quality of an optimised neutron-guide system developed for beam-tube R2 of the HOR. The HOR is a pool-type nuclear research reactor at the Interfaculty Reactor Institute of the Delft University of Technology. It is shown that the quality factors of the newly installed neutron guides are between 0.49 and 0.63.

  3. A new method to determine in situ the transmission of a neutron-guide system at a reactor source

    International Nuclear Information System (INIS)

    Haan, V.O. de; Gibcus, H.P.M.; Gommers, R.M.; Labohm, F.; Well, A.A. van; Leege, P.F.A. de; Schebetov, A.; Pusenkov, V.

    2002-01-01

    In this paper, a description of a new method to determine the transmission of neutron guides after they are installed in a beam-tube at a reactor source is given. The method is based on activation measurements of gold foils at the entrance of the beam-tube and at the exit of the neutron guides compared to Monte-Carlo calculations. In this method, a quality factor is defined as the ratio between the actual transmission and the theoretical maximum attainable transmission. This method is used to determine the quality of an optimised neutron-guide system developed for beam-tube R2 of the HOR. The HOR is a pool-type nuclear research reactor at the Interfaculty Reactor Institute of the Delft University of Technology. It is shown that the quality factors of the newly installed neutron guides are between 0.49 and 0.63

  4. 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.

  5. Fast reactor fuel reprocessing. An Indian perspective

    International Nuclear Information System (INIS)

    Natarajan, R.; Raj, Baldev

    2005-01-01

    The Department of Atomic Energy (DAE) envisioned the introduction of Plutonium fuelled fast reactors as the intermediate stage, between Pressurized Heavy Water Reactors and Thorium-Uranium-233 based reactors for the Indian Nuclear Power Programme. This necessitated the closing of the fast reactor fuel cycle with Plutonium rich fuel. Aiming to develop a Fast Reactor Fuel Reprocessing (FRFR) technology with low out of pile inventory, the DAE, with over four decades of operating experience in Thermal Reactor Fuel Reprocessing (TRFR), had set up at the India Gandhi Center for Atomic Research (IGCAR), Kalpakkam, R and D facilities for fast reactor fuel reprocessing. After two decades of R and D in all the facets, a Pilot Plant for demonstrating FRFR had been set up for reprocessing the FBTR (Fast Breeder Test Reactor) spent mixed carbide fuel. Recently in this plant, mixed carbide fuel with 100 GWd/t burnup fuel with short cooling period had been successfully reprocessed for the first time in the world. All the challenging problems encountered had been successfully overcome. This experience helped in fine tuning the designs of various equipments and processes for the future plants which are under construction and design, namely, the DFRP (Demonstration Fast reactor fuel Reprocessing Plant) and the FRP (Fast reactor fuel Reprocessing Plant). In this paper, a comprehensive review of the experiences in reprocessing the fast reactor fuel of different burnup is presented. Also a brief account of the various developmental activities and strategies for the DFRP and FRP are given. (author)

  6. A Study on Conjugate Heat Transfer Analysis of Reactor Vessel including Irradiated Structural Heat Source

    Energy Technology Data Exchange (ETDEWEB)

    Yi, Kunwoo; Cho, Hyuksu; Im, Inyoung; Kim, Eunkee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    Though Material reliability programs (MRPs) have a purpose to provide the evaluation or management methodologies for the operating RVI, the similar evaluation methodologies can be applied to the APR1400 fleet in the design stage for the evaluation of neutron irradiation effects. The purposes of this study are: to predict the thermal behavior whether or not irradiated structure heat source; to evaluate effective thermal conductivity (ETC) in relation to isotropic and anisotropic conductivity of porous media for APR1400 Reactor Vessel. The CFD simulations are performed so as to evaluate thermal behavior whether or not irradiated structure heat source and effective thermal conductivity for APR1400 Reactor Vessel. In respective of using irradiated structure heat source, the maximum temperature of fluid and core shroud for isotropic ETC are 325.8 .deg. C, 341.5 .deg. C. The total amount of irradiated structure heat source is about 5.41 MWth and not effect to fluid temperature.

  7. Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

    International Nuclear Information System (INIS)

    Potapov, I.A.; Serebrov, A.P.

    2001-01-01

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH 2 ) and liquid deuterium (LD 2 ) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

  8. Advances in automated noise data acquisition and noise source modeling for power reactors

    International Nuclear Information System (INIS)

    Clapp, N.E. Jr.; Kryter, R.C.; Sweeney, F.J.; Renier, J.A.

    1981-01-01

    A newly expanded program, directed toward achieving a better appreciation of both the strengths and limitations of on-line, noise-based, long-term surveillance programs for nuclear reactors, is described. Initial results in the complementary experimental (acquisition and automated screening of noise signatures) and theoretical (stochastic modeling of likely noise sources) areas of investigation are given

  9. Canadian Neutron Source (CNS): a research reactor solution for medical isotopes and neutrons for science

    International Nuclear Information System (INIS)

    Chapman, D.

    2009-01-01

    This presentation describes a dual purpose research facility at the University of Saskatchewan for Canada for the production of medical isotopes and neutrons for scientific research. The proposed research reactor is intended to supply most of Canada's medical isotope requirements and provide a neutron source for Canada's research community. Scientific research would include materials research, biomedical research and imaging.

  10. Regulatory Technology Development Plan Sodium Fast Reactor. Mechanistic Source Term Development

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David S. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, Acacia Joann [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew D. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-02-28

    Construction and operation of a nuclear power installation in the U.S. requires licensing by the U.S. Nuclear Regulatory Commission (NRC). A vital part of this licensing process and integrated safety assessment entails the analysis of a source term (or source terms) that represents the release of radionuclides during normal operation and accident sequences. Historically, nuclear plant source term analyses have utilized deterministic, bounding assessments of the radionuclides released to the environment. Significant advancements in technical capabilities and the knowledge state have enabled the development of more realistic analyses such that a mechanistic source term (MST) assessment is now expected to be a requirement of advanced reactor licensing. This report focuses on the state of development of an MST for a sodium fast reactor (SFR), with the intent of aiding in the process of MST definition by qualitatively identifying and characterizing the major sources and transport processes of radionuclides. Due to common design characteristics among current U.S. SFR vendor designs, a metal-fuel, pool-type SFR has been selected as the reference design for this work, with all phenomenological discussions geared toward this specific reactor configuration. This works also aims to identify the key gaps and uncertainties in the current knowledge state that must be addressed for SFR MST development. It is anticipated that this knowledge state assessment can enable the coordination of technology and analysis tool development discussions such that any knowledge gaps may be addressed. Sources of radionuclides considered in this report include releases originating both in-vessel and ex-vessel, including in-core fuel, primary sodium and cover gas cleanup systems, and spent fuel movement and handling. Transport phenomena affecting various release groups are identified and qualitatively discussed, including fuel pin and primary coolant retention, and behavior in the cover gas and

  11. Impact of source terms on distances to which reactor accident consequences occur

    International Nuclear Information System (INIS)

    Ostmeyer, R.M.

    1982-01-01

    Estimates of the distances over which reactor accident consequences might occur are important for development of siting criteria and for emergency response planning. This paper summarizes the results of a series of CRAC2 calculations performed to estimate these distances. Because of the current controversy concerning the magnitude of source terms for severe accidents, the impact of source term reductions upon distance estimates is also examined

  12. Compact radio sources as a plasma turbulent reactor

    International Nuclear Information System (INIS)

    Atoyan, A.M.; Nagapetyan, A.

    1987-01-01

    The electromagnetic raiation spectra of a homogeneous cosmic radio source (CRS) wherein the relativistic electron acceleration on the langmuir waves leads to the formation of Maxwell-like spectra with characteristic value of the Lorentz-factor γ 0 ∼ 10 3 are considered. It has been shown that due to synchrotron radiation of relativistic electrons, usually observed from CRSs flat radiosepctra, gradually steepening at submillimeter wavelengths are naturally formed in the optically thin range of frequencies. The electromagnetic radiation at the scattering of the electron on the turbulence produces significant nonthermal infrared radiation. Inverse compton scattering of the relativistic electrons on the radio-infrared photons leads the production of X-rays. The characteristic of the electromagnetic radiation spectra obtained in the model are compared with the observational ones

  13. Productivity of a nuclear chemical reactor with gamma radioisotopic sources; Rendimiento de un reactor quimico-nuclear con fuentes radioisotopicas gamma

    Energy Technology Data Exchange (ETDEWEB)

    Anguis T, C

    1975-07-01

    According to an established mathematical model of successive Compton interaction processes the made calculations for major distances are extended checking the acceptability of the spheric geometry model for the experimental data for radioisotopic sources of Co-60 and Cs-137. Parameters such as the increasing factor and the absorbed dose served as comparative base. calculations for the case of a punctual source succession inside a determined volume cylinder are made to obtain the total dose, the deposited energy by each photons energetic group and the total absorbed energy inside the reactor. Varying adequately the height/radius relation for different cylinders, the distinct energy depositions are compared in each one of them once a time standardized toward a standard value of energy emitted by the reactor volume. A relation between the quantity of deposited energy in each point of the reactor and the conversion values of chemical species is established. They are induced by electromagnetic radiation and that are reported as ''G'' in the scientific literature (number of molecules formed or disappeared by each 100 e.v. of energy). Once obtained the molecular performance inside the reactor for each type of geometry, it is optimized the height/radius relation according to the maximum production of molecules by unity of time. It is completed a bibliographical review of ''G'' values reported by different types of aqueous solutions with the purpose to determine the maximum performance of molecular hydrogen as a function of pH of the solution and of the used type of solute among other factors. Calculations for the ethyl bromide production as an example of one of the industrial processes which actually work using the gamma radiation as reactions inductor are realized. (Author)

  14. Productivity of a nuclear chemical reactor with gamma radioisotopic sources; Rendimiento de un reactor quimico-nuclear con fuentes radioisotopicas gamma

    Energy Technology Data Exchange (ETDEWEB)

    Anguis T, C

    1975-07-01

    According to an established mathematical model of successive Compton interaction processes the made calculations for major distances are extended checking the acceptability of the spheric geometry model for the experimental data for radioisotopic sources of Co-60 and Cs-137. Parameters such as the increasing factor and the absorbed dose served as comparative base. calculations for the case of a punctual source succession inside a determined volume cylinder are made to obtain the total dose, the deposited energy by each photons energetic group and the total absorbed energy inside the reactor. Varying adequately the height/radius relation for different cylinders, the distinct energy depositions are compared in each one of them once a time standardized toward a standard value of energy emitted by the reactor volume. A relation between the quantity of deposited energy in each point of the reactor and the conversion values of chemical species is established. They are induced by electromagnetic radiation and that are reported as ''G'' in the scientific literature (number of molecules formed or disappeared by each 100 e.v. of energy). Once obtained the molecular performance inside the reactor for each type of geometry, it is optimized the height/radius relation according to the maximum production of molecules by unity of time. It is completed a bibliographical review of ''G'' values reported by different types of aqueous solutions with the purpose to determine the maximum performance of molecular hydrogen as a function of pH of the solution and of the used type of solute among other factors. Calculations for the ethyl bromide production as an example of one of the industrial processes which actually work using the gamma radiation as reactions inductor are realized. (Author)

  15. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele [Mainz Univ. (Germany). Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren (AFR)

    2012-10-15

    AFR (Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren) is the German acronym for 'Association for Research Reactor Operation and Safety Issues' which was founded in 1959. Reactor managers of European research reactors mainly from the German linguistic area meet regularly for their mutual benefit to exchange experience and knowledge in all areas of operating, managing and utilization of research reactors. In the last 2 years joint meetings were held together with the French association of research reactors CER (Club d'Exploitants des Reacteurs). In this contribution the AFR, its members, work and aims as well as the French partner CER are presented. (orig.)

  16. INR TRIGA Research Reactors: A Neutron Source for Radioisotopes and Materials Investigation

    International Nuclear Information System (INIS)

    Barbos, D.; Ciocanescu, M.; Paunoiu, C.; Bucsa, A.F.

    2013-01-01

    At the INR there are 2 high intensity neutron sources. These sources are in fact the two nuclear TRIGA reactors: TRIGA SSR 14 MW and TRIGA ACPR. TRIGA stationary reactor is provided with several in-core irradiation channels. Other several out-of-core irradiation channels are located in the vertical channels in the beryllium reflector blocks. The maximum value of the thermal neutron flux (E 14 cm -2 s -1 and of fast neutron flux (E>1 MeV) is 6.89×10 13 cm -2 s -1 . For neutron activation analysis both reactors are used and k0-NAA method has been implemented. At INR Pitesti a prompt gamma ray neutron activation analysis devices has been designed, manufactured ant put into operation. For nuclear materials properties investigation neutron radiography methods was developed in INR. For these purposes two neutron radiography devices were manufacture, one of them underwater and other one dry. The neutron beams are used for investigation of materials properties and components produced or under development for applications in the energy sector (fission and fusion). At TRIGA 14 MW reactor a neutron difractormeter and a SANS devices are available for material residual stress and texture measurements. TRIGA 14 MW reactor is used for medical and industrial radioisotopes production ( 131 I, 125 I, 192 Ir, etc) and a method for 99 Mo- 99 Tc production from fission is under developing. At INR Pitesti several special programmes for new types of nuclear fuel behavior characterization are under development. (author)

  17. Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

    2016-04-17

    In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

  18. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    International Nuclear Information System (INIS)

    Son, Han Seong; Song, Deok Yong; Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon

    2006-01-01

    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors

  19. Optimization study of ultracold neutron sources at TRIGA reactors using MCNP

    International Nuclear Information System (INIS)

    Pokotilovskij, Yu.N.; Rogov, A.D.

    1997-01-01

    Monte Carlo simulation for the optimization of ultracold and very cold neutron sources for TRIGA reactors is performed. The calculations of thermal and cold neutron fluxes from the TRIGA reactor for different positions and configurations of a very cold solid methane moderator were performed with using the MCNP program. The production of neutrons in the ultracold and very cold energy range was calculated for the most promising final moderators (converters): very cold solid deuterium and heavy methane. The radiation energy deposition was calculated for the optimized solid methane-heavy methane cold neutron moderator

  20. 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)

  1. EPR by AREVA. An evolutionary reactor

    International Nuclear Information System (INIS)

    Horstmann, Marion

    2010-01-01

    The EPR development goals are as follows: 1. Evolutionary design to fully capitalize on the design, construction and operating experience based on the 86 AREVA's PWR operating worldwide; 2. Enhanced Safety compared to operating PWRs: reduce core damage frequency (CDF), accommodate severe accidents with no long-term population effect, Withstand large airplane crash (APC); 3. High availability; 4. Simplified operation and maintenance; and 5. Generation cost at least 10 % lower than 1500 MWe series in operation.The design builds on the achievements of the N4 and Konvoi reactors. The main plant data are tabulated. The PWR structure is shown as an example of the stepwise improvement. Focus of the presentation is on the construction techniques, supply chain, and project delivery. (P.A.)

  2. The GBR reactor an economically competitive Breeder

    International Nuclear Information System (INIS)

    Chermanne, J.

    1974-01-01

    In this article the design is described of a 1200 MWe fast breeder, gas-cooled reactor (GBR-4), prepared by a group of experts of the Gas Breeder Reactor Association and used as a reference system for economical and safety evaluations, as well as for defining the research and development program focussed on such concept and the specifications of the prospective demonstrative plant

  3. Experiment on search for neutron-antineutron oscillations using a projected UCN source at the WWR-M reactor

    Science.gov (United States)

    Fomin, A. K.; Serebrov, A. P.; Zherebtsov, O. M.; Leonova, E. N.; Chaikovskii, M. E.

    2017-01-01

    We propose an experiment on search for neutron-antineutron oscillations based on the storage of ultracold neutrons (UCN) in a material trap. The sensitivity of the experiment mostly depends on the trap size and the amount of UCN in it. In Petersburg Nuclear Physics Institute (PNPI) a high-intensity UCN source is projected at the WWR-M reactor, which must provide UCN density 2-3 orders of magnitude higher than existing sources. The results of simulations of the designed experimental scheme show that the sensitivity can be increased by ˜ 10-40 times compared to sensitivity of previous experiment depending on the model of neutron reflection from walls.

  4. Calculation Of Fuel Burnup And Radionuclide Inventory In The Syrian Miniature Neutron Source Reactor Using The GETERA Code

    International Nuclear Information System (INIS)

    Khattab, K.; Dawahra, S.

    2011-01-01

    Calculations of the fuel burnup and radionuclide inventory in the Syrian Miniature Neutron Source Reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer and has a bigger library of isotopes and more accurate. (author)

  5. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Ruck, G.W.; Lee, A.Y.; Smeltzer, G.; Prevenslik, T.

    1978-11-01

    A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

  6. Source-term reevaluation for US commercial nuclear power reactors: a status report

    International Nuclear Information System (INIS)

    Herzenberg, C.L.; Ball, J.R.; Ramaswami, D.

    1984-12-01

    Only results that had been discussed publicly, had been published in the open literature, or were available in preliminary reports as of September 30, 1984, are included here. More than 20 organizations are participating in source-term programs, which have been undertaken to examine severe accident phenomena in light-water power reactors (including the chemical and physical behavior of fission products under accident conditions), update and reevaluate source terms, and resolve differences between predictions and observations of radiation releases and related phenomena. Results from these source-term activities have been documented in over 100 publications to date

  7. Application of the modified neutron source multiplication method for a measurement of sub-criticality in AGN-201K reactor

    International Nuclear Information System (INIS)

    Myung-Hyun Kim

    2010-01-01

    Measurement of sub-criticality is a challenging and required task in nuclear industry both for nuclear criticality safety and physics test in nuclear power plant. A relatively new method named as Modified Neutron Source Multiplication Method (MNSM) was proposed in Japan. This method is an improvement of traditional Neutron Source Multiplication (NSM) Method, in which three correction factors are applied additionally. In this study, MNSM was tested in calculation of rod worth using an educational reactor in Kyung Hee University, AGN-201K. For this study, a revised nuclear data library and a neutron transport code system TRANSX-PARTISN were used for the calculation of correction factors for various control rod positions and source locations. Experiments were designed and performed to enhance errors in NSM from the location effects of source and detectors. MNSM can correct these effects but current results showed not much correction effects. (author)

  8. An 'open source' networked identity

    DEFF Research Database (Denmark)

    Larsen, Malene Charlotte

    2013-01-01

    This paper examines the concept of identity in relation to youth practices on social network sites (SNS). The paper illustrates how writing “I love you” or other emotional statements on each other’s profiles on SNS is not only a common way for Danish teenagers to communicate and practice friendship...... communicative actions – are not only performing their own identity, but are becoming co-constructors of each other's identities, which the author characterizes as an 'open source' networked identity....

  9. Uncertainties in source term calculations generated by the ORIGEN2 computer code for Hanford Production Reactors

    International Nuclear Information System (INIS)

    Heeb, C.M.

    1991-03-01

    The ORIGEN2 computer code is the primary calculational tool for computing isotopic source terms for the Hanford Environmental Dose Reconstruction (HEDR) Project. The ORIGEN2 code computes the amounts of radionuclides that are created or remain in spent nuclear fuel after neutron irradiation and radioactive decay have occurred as a result of nuclear reactor operation. ORIGEN2 was chosen as the primary code for these calculations because it is widely used and accepted by the nuclear industry, both in the United States and the rest of the world. Its comprehensive library of over 1,600 nuclides includes any possible isotope of interest to the HEDR Project. It is important to evaluate the uncertainties expected from use of ORIGEN2 in the HEDR Project because these uncertainties may have a pivotal impact on the final accuracy and credibility of the results of the project. There are three primary sources of uncertainty in an ORIGEN2 calculation: basic nuclear data uncertainty in neutron cross sections, radioactive decay constants, energy per fission, and fission product yields; calculational uncertainty due to input data; and code uncertainties (i.e., numerical approximations, and neutron spectrum-averaged cross-section values from the code library). 15 refs., 5 figs., 5 tabs

  10. Upgrade of the ultracold neutron source at the pulsed reactor TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Kahlenberg, J.; Ross, K.U.; Beck, M.; Heil, W.; Karch, J.; Kories, F.; Kretschmer, M. [Johannes Gutenberg University, Institute of Physics, Mainz (Germany); Ries, D. [Johannes Gutenberg University, Institute of Nuclear Chemistry, Mainz (Germany); Paul Scherrer Institute (PSI), Laboratory for Particle Physics, Villigen (Switzerland); ETH Zuerich, Institute for Particle Physics, Zuerich (Switzerland); Siemensen, C.; Geppert, C.; Karpuk, S.; Reich, T.; Sobolev, Y.; Trautmann, N. [Johannes Gutenberg University, Institute of Nuclear Chemistry, Mainz (Germany); Hild, N. [Paul Scherrer Institute (PSI), Laboratory for Particle Physics, Villigen (Switzerland); ETH Zuerich, Institute for Particle Physics, Zuerich (Switzerland); Lauss, B. [Paul Scherrer Institute (PSI), Laboratory for Particle Physics, Villigen (Switzerland)

    2017-11-15

    The performance of the upgraded solid deuterium ultracold neutron source at the pulsed reactor TRIGA Mainz is described. The current configuration stage comprises the installation of a He liquefier to run UCN experiments over long-term periods, the use of stainless steel neutron guides with improved transmission as well as sputter-coated non-magnetic {sup 58}NiMo alloy at the inside walls of the thermal bridge and the converter cup. The UCN yield was measured in a ''standard'' UCN storage bottle (stainless steel) with a volume of 32 litres outside the biological shield at the experimental area yielding UCN densities of 8.5/cm{sup 3}; an increase by a factor of 3.5 compared to the former setup. The measured UCN storage curve is in good agreement with the predictions from a Monte Carlo simulation developed to model the source. The growth and formation of the solid deuterium converter during freeze-out are affected by the ortho/para ratio of the H{sub 2} premoderator. (orig.)

  11. Optimization by simulation of the coupling between a sub-critical reactor and its spallation source. Towards a pilot reactor

    International Nuclear Information System (INIS)

    Kerdraon, D.

    2001-10-01

    Accelerator Driven Systems (ADS), based on a proton accelerator and a sub-critical core coupled with a spallation target, offer advantages in order to reduce the nuclear waste radiotoxicity before repository closure. Many studies carried out on the ADS should lead to the definition of an experimental plan which would federate the different works in progress. This thesis deals with the neutronic Monte Carlo simulations with the MCNPX code to optimize such a system in view of a pilot reactor building. First, we have recalled the main neutronic properties of an hybrid reactor. The concept of gas-cooled eXperimental Accelerator Driven System (XADS) chosen for our investigations comes from the preliminary studies done by the Framatome company. In order to transmute minor actinides, we have considered the time evolution of the main fuels which could be reasonably used for the demonstration phases. The neutronic parameters of the reactor, concerning minor actinide transmutation, are reported. Also, we have calculated the characteristic times and the transmutation rates in the case of 99 Tc and 129 I isotopes. We have identified some neutronic differences between an experimental and a power ADS according to the infinite multiplication coefficient, the shape factor and the level of flux to extend the demonstrator concept. We have proposed geometric solutions to keep the radial shape factor of a power ADS acceptable. In the last part, beyond the experimental XADS scope, we have examined the possible transition towards an uranium/thorium cycle based on Molten Salt Reactors using a power ADS in order to generate the required 233 U proportion. (author)

  12. Plant protection system optimization studies to mitigate consequences of large breaks in the advanced neutron source reactor

    International Nuclear Information System (INIS)

    Khayat, M.I.; March-Leuba, J.

    1993-01-01

    This paper documents some of the optimization studies performed to maximize the performance of the engineered safety features and scram systems to mitigate the consequences of large breaks in the primary cooling system of the advanced neutron source (ANS) reactor. The ANS is a new basic and applied research facility based on a powerful steady-state research reactor that provides beams of neutrons for measurements and experiments in the field of material science and engineering, biology, chemistry, material analysis, and nuclear science. To achieve the high neutron fluxes for these state-of-the-art experiments, the ANS design has a very high power density core (330 MW fission with an active volume of 67.6 ell) surrounded by a large heavy-water reflector, where most neutrons are moderated. This design maximizes the number of neutrons available for experiments but results in a low heat capacity core that creates unique challenges to the design of the plant protection system

  13. Radioisotope tracer study in an aniline production reactor

    International Nuclear Information System (INIS)

    Pant, H.J.; Yelgoankar, V.N.; Mendhekar, G.N.

    1995-01-01

    A radioisotope tracer study was carried out in an aniline production reactor to investigate the cause of poor heat transfer from tube side to shell side in an aniline production (ANPO) reactor. The results of the study indicated that more than 50% of the shell volume was reduced due to deposition of the process material (i.e. fouling) on the shell walls and may be the cause of poor heat transfer in the reactor. (author). 2 refs., 2 figs

  14. Performance of the advanced cold neutron source and optics upgrades at the NIST Research Reactor

    International Nuclear Information System (INIS)

    Williams, R.E.; Kopetka, P.; Cook, J.C.; Rowe, J.M.

    2003-01-01

    On March 6, 2002, the NIST Research Reactor resumed routine operation following a six-month shutdown for facility upgrades and maintenance. During the shutdown, the original liquid hydrogen cold neutron source was removed, and the advanced cold source was installed. An optical filter was installed on one of the neutron guides, NG-3, replacing a crystal filter for the 30-m SANS instrument and the guide used between the chopper disks of the Disk Chopper time-of-flight Spectrometer (DCS) installed on NG-4 has been recently reconfigured. Additional improvements in the neutron optics of various instruments are being made. The advanced liquid hydrogen cold neutron source performs as expected, nearly doubling the flux available to most instruments. The measured gains range from about 1.4 at 2 A, to over a factor of two at 15 A. Also as expected, the heat load in the new source increased to 1200 watts, but the previously existing refrigerator has easily accommodated the increase. With intensity gains of a factor of two in the important long wavelength region of the spectrum, the advanced cold source significantly enhances the measurement capability of the cold neutron scattering instrumentation at NIST. The optical filter on NG-3 is also very successful; the 30-m SANS has an additional gain of two at 17 A. A system of refracting lenses and prisms near the SANS sample position has made possible measurements at low Q (0.0005 A -1 ) that were previously not feasible. The DCS has also seen additional intensity gain factors in excess of two for the majority of experiments and at short neutron wavelengths the gains exceed three. In addition, two new triple axis spectrometers will feature double-focusing monochromators in order to exploit the full size of the available thermal and cold neutron beam tubes. The success of the advanced cold source and enhanced neutron optics contributed to the recognition of the NIST Center for Neutron Research as 'the premiere neutron scattering

  15. Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

  16. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given

  17. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given. (author)

  18. Fuel, structural material and coolant for an advanced fast micro-reactor

    International Nuclear Information System (INIS)

    Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

    2011-01-01

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

  19. Regulatory impact of nuclear reactor accident source term assumptions. Technical report

    International Nuclear Information System (INIS)

    Pasedag, W.F.; Blond, R.M.; Jankowski, M.W.

    1981-06-01

    This report addresses the reactor accident source term implications on accident evaluations, regulations and regulatory requirements, engineered safety features, emergency planning, probabilistic risk assessment, and licensing practice. Assessment of the impact of source term modifications and evaluation of the effects in Design Basis Accident analyses, assuming a change of the chemical form of iodine from elemental to cesium iodide, has been provided. Engineered safety features used in current LWR designs are found to be effective for all postulated combinations of iodine source terms under DBA conditions. In terms of potential accident consequences, it is not expected that the difference in chemical form between elemental iodine and cesium iodide would be significant. In order to account for the current information on source terms, a spectrum of accident scenerios is discussed to realistically estimate the source terms resulting from a range of potential accident conditions

  20. An advanced method of heterogeneous reactor theory

    International Nuclear Information System (INIS)

    Kochurov, B.P.

    1994-08-01

    Recent approaches to heterogeneous reactor theory for numerical applications were presented in the course of 8 lectures given in JAERI. The limitations of initial theory known after the First Conference on Peacefull Uses of Atomic Energy held in Geneva in 1955 as Galanine-Feinberg heterogeneous theory:-matrix from of equations, -lack of consistent theory for heterogeneous parameters for reactor cell, -were overcome by a transformation of heterogeneous reactor equations to a difference form and by a development of a consistent theory for the characteristics of a reactor cell based on detailed space-energy calculations. General few group (G-number of groups) heterogeneous reactor equations in dipole approximation are formulated with the extension of two-dimensional problem to three-dimensions by finite Furie expansion of axial dependence of neutron fluxes. A transformation of initial matrix reactor equations to a difference form is presented. The methods for calculation of heterogeneous reactor cell characteristics giving the relation between vector-flux and vector-current on a cell boundary are based on a set of detailed space-energy neutron flux distribution calculations with zero current across cell boundary and G calculations with linearly independent currents across the cell boundary. The equations for reaction rate matrices are formulated. Specific methods were developed for description of neutron migration in axial and radial directions. The methods for resonance level's approach for numerous high-energy resonances. On the basis of these approaches the theory, methods and computer codes were developed for 3D space-time react or problems including simulation of slow processes with fuel burn-up, control rod movements, Xe poisoning and fast transients depending on prompt and delayed neutrons. As a result reactors with several thousands of channels having non-uniform axial structure can be feasibly treated. (author)

  1. Sources of radioactive waste from light-water reactors and their physical and chemical properties

    International Nuclear Information System (INIS)

    Bell, M.J.; Collins, J.T.

    1979-01-01

    The general physical and chemical properties of waste streams in light-water reactors (LWRs) are described. The principal mechanisms for release and the release pathways to the environment are discussed. The calculation of liquid and gaseous source terms using one of the available models is presented. These calculated releases are compared with observed releases from operating LWRs. The computerized mathematical model used is the GALE Code which is the Nuclear Regulatory Commission (NRC) staff's model for calculating source terms for effluents from LWRs (USNRC76a, USNRC76b). Programs currently being conducted at operating reactors by the NRC, Electric Power Research Institute, and various utilities to better define the characteristics of waste streams and the performance of radwaste process equipment are described

  2. Reactor-building-basement radionuclide and source distribution studies. Volume 3

    International Nuclear Information System (INIS)

    Cox, T.E.; Horan, J.T.; Worku, G.

    1983-06-01

    The Three Mile Island Unit 2 (TMI-2) Reactor Building basement has been sampled several times since August 1979. This report compiles the analytical results and sample history for the liquid and solid samples obtained to date. In addition, basement radiation levels were also obtained using thermoluminescent dosimeters (TLDs). The data obtained will provide information to support ongoing mass balance and source term studies and will aid in characterizing the 282-ft elevation for decontamination planning and dose reduction

  3. PRISM: An innovative liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

    1986-01-01

    This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design

  4. A Liquid Deuterium Cold Neutron Source for the NIST Research Reactor - Conceptual Design

    International Nuclear Information System (INIS)

    Williams, R. E.; Middleton, M.; Kopetka, P.; Rowe, J. M.; Brand, P. C.

    2013-01-01

    The NBSR is a 20 MW research reactor operated by the NIST Center for Neutron Research (NCNR) as a neutron source providing beams of thermal and cold neutrons for research in materials science, fundamental physics and nuclear chemistry. A large, 550 mm diameter beam port was included in the design for the installation of a cold neutron source, and the NCNR has been steadily improving its cold neutron facilities for more than 25 years. Monte Carlo Simulations have shown that a liquid deuterium (LD 2 ) source will provide a gain of 1.5 to 2 for neutron wavelengths between 4 A and 10 A with respect to the existing liquid hydrogen cold source. The conceptual design for the LD 2 source will be presented. To achieve these gains, a large volume (35 litres) of LD 2 is required. The expected nuclear heat load in this moderator and vessel is 4000 W. A new, 7 kW helium refrigerator is being built to provide the necessary cooling capacity; it will be completely installed and tested early in 2014. The source will operate as a naturally circulating thermosiphon, very similar to the horizontal cold source in the High Flux Reactor at the Institut Laue-Langevin (ILL) in Grenoble. A condenser will be mounted on the reactor face about 2 m above the source providing the gravitational head to supply the source with LD 2 . The system will always be open to a 16 m3 ballast tank to store the deuterium at 500 kPa when the refrigerator is not operating, and providing a passively safe response to a refrigerator trip. It is expected the source will operate at 23 K, the boiling point of LD 2 at 100 kPa. All components will be surrounded by a blanket of helium to prevent the possibility of creating a flammable mixture of deuterium and air. A design for the cryostat assembly, consisting of the moderator chamber, vacuum jacket, helium containment and a heavy water cooling water jacket, has been completed and sent to procurement to solicit bids. It is expected that installation of the LD 2 cold

  5. A Liquid Deuterium Cold Neutron Source for the NIST Research Reactor - Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R. E.; Middleton, M.; Kopetka, P.; Rowe, J. M.; Brand, P. C. [NIST Center for Neutron Research, Gaithersburg (United States)

    2013-07-01

    The NBSR is a 20 MW research reactor operated by the NIST Center for Neutron Research (NCNR) as a neutron source providing beams of thermal and cold neutrons for research in materials science, fundamental physics and nuclear chemistry. A large, 550 mm diameter beam port was included in the design for the installation of a cold neutron source, and the NCNR has been steadily improving its cold neutron facilities for more than 25 years. Monte Carlo Simulations have shown that a liquid deuterium (LD{sub 2}) source will provide a gain of 1.5 to 2 for neutron wavelengths between 4 A and 10 A with respect to the existing liquid hydrogen cold source. The conceptual design for the LD{sub 2} source will be presented. To achieve these gains, a large volume (35 litres) of LD{sub 2} is required. The expected nuclear heat load in this moderator and vessel is 4000 W. A new, 7 kW helium refrigerator is being built to provide the necessary cooling capacity; it will be completely installed and tested early in 2014. The source will operate as a naturally circulating thermosiphon, very similar to the horizontal cold source in the High Flux Reactor at the Institut Laue-Langevin (ILL) in Grenoble. A condenser will be mounted on the reactor face about 2 m above the source providing the gravitational head to supply the source with LD{sub 2}. The system will always be open to a 16 m3 ballast tank to store the deuterium at 500 kPa when the refrigerator is not operating, and providing a passively safe response to a refrigerator trip. It is expected the source will operate at 23 K, the boiling point of LD{sub 2} at 100 kPa. All components will be surrounded by a blanket of helium to prevent the possibility of creating a flammable mixture of deuterium and air. A design for the cryostat assembly, consisting of the moderator chamber, vacuum jacket, helium containment and a heavy water cooling water jacket, has been completed and sent to procurement to solicit bids. It is expected that

  6. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

    1997-01-01

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

  7. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  8. An overview-probabilistic safety analysis for research reactors

    International Nuclear Information System (INIS)

    Liu Jinlin; Peng Changhong

    2015-01-01

    For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

  9. Small liquid metal reactor for an initial phase of fast breeder reactor introduction

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1985-01-01

    Safety and burnup characteristics of a 1000 MWth liquid metal reactor have been examined for various fuel types. With metallic Pu/Th fuel containing a small amount of zirconium hydride, low sodium-void reactivity, a high Doppler coefficient, and small burnup reactivity swings can be achieved. A conservative design is considered for an initial phase of fast breeder reactor development and possible modifications are discussed. (Author) [pt

  10. An abnormal event advisory expert system prototype for reactor operators

    International Nuclear Information System (INIS)

    Hance, D.C.

    1989-01-01

    Nuclear plant operators must respond correctly during abnormal conditions in the presence of dynamic and potentially overwhelming volumes of information. For this reason, considerable effort has been directed toward the development of nuclear plant operator aids using artificial intelligence techniques. The objective of such systems is to diagnose abnormal conditions within the plant, possibly predict consequences, and advise the operators of corrective actions in a timely manner. The objective of the work is the development of a prototype expert system to diagnose abnormal events at a nuclear power plant and advise plant operators of the event and applicable procedures in an on-line mode. The major difference between this effort and previous work is the use of plant operating procedures as a knowledge source and as an integral part of the advice provided by the expert system. The acceptance by utilities of expert systems as operator aids requires that such systems be compatible with the regulatory environment and provide economic benefits. For this reason, commercially viable operator aid systems developed in the near future must complement existing plant procedures rather than reach beyond them in a revolutionary manner. A knowledge source is the resource providing facts and relationships that are coded into the expert system program. In this case, the primary source of knowledge is a set of selected abnormal operating procedures for a modern Westinghouse pressurized water reactor

  11. An inherently safe power reactor module

    International Nuclear Information System (INIS)

    Salerno, L.N.

    1985-01-01

    General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

  12. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1978-01-01

    This law has following two purposes. At first, it exercises necessary controls concerning nuclear source material, nuclear fuel material and reactors in order to: (a) limit their uses to those for the peaceful purpose; (b) ensure planned uses of them; and (c) ensure the public safety by preventing accidents from their uses. Necessary controls are to be made concerning the refining, fabricating and reprocessing businesses, as well as the construction and operation of reactors. The second purpose of the law is to exercise necessary controls concerning internationally controlled material in order to execute the treaties and other international agreements on the research, development and use of atomic energy (the first chapter). In the second and following chapters the law prescribes controls for the persons who wish to carry on the refining and fabricating businesses, to construct and operate reactors, and to conduct the reprocessing business, as well as for those who use the internationally controlled material, respectively in separate chapters by the category of those businesses. For example, the controls to the person who wishes to construct and operate reactors are: (a) the permission of the business after the examination; (b) the examination and approval of the design and methods of construction prior to the construction; (c) the inspection of the facilities prior to their use; (d) periodic inspections of the facilities; (e) the establishment of requirements for safety measures and punishments to their violations. (Matsushima, A.)

  13. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications

    International Nuclear Information System (INIS)

    Carluccio, Thiago

    2011-01-01

    This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as k eff and k src , and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

  14. Nuclear reactor having an inflatable vessel closure seal structure

    International Nuclear Information System (INIS)

    1980-01-01

    An improved type of closure head seal for the rotatable plugs of the reactor vessel of a liquid metal fast breeder reactor is described. The seal prevents the release of radioactive particles while allowing the plug to be rotated without major manipulation of the seal structure. (UK)

  15. An approach to neutronics analysis of candu reactors

    International Nuclear Information System (INIS)

    Gul, S.; Arshad, M.

    1982-12-01

    An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Until now CANDU reactors have been analysed by the methods developed at AECL and CGE using mainly receipe methods. Relying on multigroup transport codes GAM-GATHER in combination with diffusion code CITATION a package of codes is established to use it for survey as well as production purposes. (authors)

  16. An advanced fusion neutron source facility

    International Nuclear Information System (INIS)

    Smith, D.L.

    1992-01-01

    Accelerator-based 14-MeV-neutron sources based on modifications of the original Fusion Materials Irradiation Facility are currently under consideration for investigating the effects of high-fluence high-energy neutron irradiation on fusion-reactor materials. One such concept for a D-Li neutron source is based on recent advances in accelerator technology associated with the Continuous Wave Deuterium Demonstrator accelerator under construction at Argonne National Laboratory, associated superconducting technology, and advances in liquid-metal technology. In this paper a summary of conceptual design aspects based on improvements in technologies is presented

  17. Study on recriticality of fuel debris during hypothetical severe accidents in the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.; Shin, S.T.

    1995-09-01

    A study has been performed to measure the potential of recriticality during hypothetical severe accident in Advanced Neutron Source (ANS). For the lumped debris configuration in the Reactor Coolant System (RCS), as found in the previous study, recriticality potential may be very low. However, if fuel debris is dispersed and mixed with heavy water in RCS, recriticality potential has been predicted to be substantial depending on thermal-hydraulic conditions surrounding fuel debris mixture. The recriticality potential in RCS is substantially reduced for the three element core design with 50% enrichment. Also, as observed in the previous study, strong dependencies of k eff on key thermal hydraulic parameters are shown. Light water contamination is shown to provide a positive reactivity, and void formation due to boiling of mixed water provides enough negative reactivity and to bring the system down to subcritical. For criticality potential in the subpile room, the lumped debris configuration does not pose a concern. Dispersed configuration in light water pool of the subpile room is also unlikely to result in criticality. However, if the debris is dispersed in the pool that is mixed with heavy water, the results indicate that a substantial potential exists for the debris to reach the criticality. However, if prompt recriticality disperses the debris completely in the subpile room pool, subsequent recriticality may be prevented since neutron leakage effects become large enough

  18. A Study on the demands of research reactors and considerations for an export

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, Young Jun

    2008-11-15

    Among around 240 research reactors in operation over the world, around 80% have been operated for more than 20 years and 65% for more than 30 years. Hence the number of operable reactors is expected, between 2010 and 2020, to be reduced to 1/3 of the present situation if the lifetime of a research reactor is assumed to be 40 years. However, considering the recent re-highlighting of nuclear energy as a practical mass energy source and the contributions to the overall areas of science and technology, the demands for constructing a new research reactor and replacing the existing research reactors will be increased in the near future. On the other hand, vendors which participate in providing research reactors are not few, and AREVA in France and INVAP in Argentina are example of them in a positive position. Japan and Russia are regarded as potential competitors, but they do not actively appear in the market so far. Comparing those competitors with Korea, we have weak points regarding experiences on exports and the organizational systems as an integrated vendor. But we may have a competitiveness by grafting our experiences on the development of nuclear power technology and the construction and operation of the HANARO. In this report, the future potential demands for research reactors and the related considerations for exports have been surveyed and described, particularly, centering around the Netherlands, Vietnam and Thailand that are countries which may construct research reactors in the near future. Considerations for exporting a research reactor have been categorized into two groups of technical and nontechnical items. From a technical point of view, the issues on fuel and reactor type, design data and design ability, design codes, and technology property rights have been reviewed. For the non-technical items, an integrated project system, reasonable estimate of demands, social and economic conditions for potential demand countries, MOU status, nuclear non

  19. A Study on the demands of research reactors and considerations for an export

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, Young Jun

    2008-11-01

    Among around 240 research reactors in operation over the world, around 80% have been operated for more than 20 years and 65% for more than 30 years. Hence the number of operable reactors is expected, between 2010 and 2020, to be reduced to 1/3 of the present situation if the lifetime of a research reactor is assumed to be 40 years. However, considering the recent re-highlighting of nuclear energy as a practical mass energy source and the contributions to the overall areas of science and technology, the demands for constructing a new research reactor and replacing the existing research reactors will be increased in the near future. On the other hand, vendors which participate in providing research reactors are not few, and AREVA in France and INVAP in Argentina are example of them in a positive position. Japan and Russia are regarded as potential competitors, but they do not actively appear in the market so far. Comparing those competitors with Korea, we have weak points regarding experiences on exports and the organizational systems as an integrated vendor. But we may have a competitiveness by grafting our experiences on the development of nuclear power technology and the construction and operation of the HANARO. In this report, the future potential demands for research reactors and the related considerations for exports have been surveyed and described, particularly, centering around the Netherlands, Vietnam and Thailand that are countries which may construct research reactors in the near future. Considerations for exporting a research reactor have been categorized into two groups of technical and nontechnical items. From a technical point of view, the issues on fuel and reactor type, design data and design ability, design codes, and technology property rights have been reviewed. For the non-technical items, an integrated project system, reasonable estimate of demands, social and economic conditions for potential demand countries, MOU status, nuclear non

  20. Analysis of an accelerator-driven subcritical light water reactor

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

    1997-01-01

    An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

  1. An overview of reactor physics standards: Past, present and future

    International Nuclear Information System (INIS)

    Cokinos, D.M.

    1992-07-01

    This report discusses for determining key static reactor physics parameters which have been developed by groups of experts (working groups) under the aegis of ANS-19, the ANS Reactor Physics Standards Committee. Following a series of sequential reviews, augmented by feedback from potential users, a proposed standard is brought into final form by the working group before it is adopted as a formal standard by the American National Standards Institute (ANSI); Reactor Physics standards are intended to provide guidance in the performance and qualification of complex sequences of reactor calculations and/or measurements and are regularly reviewed for possible updates and/or revisions. The reactor physics standards developed to date are listed and standards now being developed by the respective working groups are also provided

  2. Introducing an ILS methodology into research reactors

    International Nuclear Information System (INIS)

    Lorenzo, N. de; Borsani, R.C.

    2003-01-01

    Integrated Logistics Support (ILS) is the managerial organisation that co-ordinates the activities of many disciplines to develop the supporting resources (training, staffing, designing aids, equipment removal routes, etc) required by technologically complex systems. The application of an ILS methodology in defence projects is described in several places, but it is infrequently illustrated for other areas; therefore the present paper deals with applying this approach to research reactors under design or already in operation. Although better results are obtained when applied since the very beginning of a project, it can be applied successfully in facilities already in operation to improve their capability in a cost-effective way. In applying this methodology, the key objectives shall be previously identified in order to tailor the whole approach. Generally in high power multipurpose reactors, obtaining maximum profit at the lowest possible cost without reducing the safety levels are key issues, while in others the goal is to minimise drawbacks like spurious shutdowns, low quality experimental results or even to reduce staff dose to ALARA values. These items need to be quantified for establishing a system status base line in order to trace the process evolution. Thereafter, specific logistics analyses should be performed in the different areas composing the system. RAMS (Reliability, Availability, Maintainability and Supportability), Manning, Training Needs, Supplying Needs are some examples of these special logistic assessments. The following paragraphs summarise the different areas, encompassed by this ILS methodology. Plant design is influenced focussing the designers? attention on the objectives already identified. Careful design reviews are performed only in an early design stage, being useless a later application. In this paper is presented a methodology including appropriate tools for ensuring the designers abide to ILS issues and key objectives through the

  3. EL-3 dismantling of an experimental reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

  4. Nuclear fission sustainability with subcritical reactors driven by external neutron sources

    International Nuclear Information System (INIS)

    Lafuente, A.; Piera, M.

    2011-01-01

    Although nuclear breeder reactors are a promising way to enhance the potential energy currently retrievable from the Uranium reserves, they still have disadvantages because of their safety features (i.e. poor stabilizing mechanisms) and the security of their fuel cycle (diversion of Pu for non-civilian purposes). Loading natural nuclear fuels to a reactor and completely burning them without reprocessing would be ideal, however, this is not possible in critical reactors due to the limitations imposed by the maximum achievable burn-up. An alternative option to attain very high percentages of nuclear natural materials exploitation, while meeting other objectives of Nuclear Sustainability, could consist of using externally-driven subcritical reactors to reach the desired high burn-ups (of the order of 30% and more) without reprocessing. Such scheme would lead to an efficient exploitation of the available raw material, without any risk of proliferation. Exploring this type of reactor concept, this paper analyzes the different ways to accomplish this goal while identifying potential setbacks.

  5. Reactor refurbishment in an outage environment

    International Nuclear Information System (INIS)

    Gowthorpe, P.; Hoare, R.

    2012-01-01

    Reactor life extension has typically been performed during specific refurbishment outages. These outages are long and costly due to the sheer complexity of the scope, not to mention the ever present discovery work. A scope of this size requires a huge labour force to execute, which poses significant challenges. The work is difficult to staff with qualified people able to execute the work smoothly and managing the required labour pool problematic. Cost and time overruns are inevitable in that environment. Reducing the cost and schedule is critical to the long term viability of reactor refurbishment projects. With planning, the total cost of the refurbishment can be reduced by managing the inspection and repairs during normal outages. Identifying what activities need to be done each outage for the life of the reactor and bringing the latest technology can make this viable. Tightly planned outages with a small well trained labour force will go a long way to reducing costs. The suite of services and tooling available to the utilities to manage their reactor integrity has improved significantly in recent years and continues to evolve. New feeder inspection technologies can provide improved inspection results for the complex feeder geometry. These improvements lead to more accurate wear rates and better predictions of component life. Feeders that need replacement based on improved inspection techniques can be replaced systematically during regular outages rather than specific refurbishment outages. Targeting areas rather than entire feeders reduces time, dose and cost. In cases where feeder replacement isn't feasible or where unpredicted wear is found, a feeder weld overlay process can be used. To manage the reactor work, new data systems are under development that allow for effective tracking of each activity performed and outcomes in a single package. (author)

  6. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  7. Modular High Temperature Gas-Cooled Reactor heat source for coal conversion

    International Nuclear Information System (INIS)

    Schleicher, R.W. Jr.; Lewis, A.C.

    1992-09-01

    In the industrial nations, transportable fuels in the form of natural gas and petroleum derivatives constitute a primary energy source nearly equivalent to that consumed for generating electric power. Nations with large coal deposits have the option of coal conversion to meet their transportable fuel demands. But these processes themselves consume huge amounts of energy and produce undesirable combustion by-products. Therefore, this represents a major opportunity to apply nuclear energy for both the environmental and energy conservation reasons. Because the most desirable coal conversion processes take place at 800 degree C or higher, only the High Temperature Gas-Cooled Reactors (HTGRs) have the potential to be adapted to coal conversion processes. This report provides a discussion of this utilization of HTGR reactors

  8. Kinetic calculations for miniature neutron source reactor using analytical and numerical techniques

    International Nuclear Information System (INIS)

    Ampomah-Amoako, E.

    2008-06-01

    The analytical methods, step change in reactivity and ramp change in reactivity as well as numerical methods, fixed point iteration and Runge Kutta-gill were used to simulate the initial build up of neutrons in a miniature neutron source reactor with and without temperature feedback effect. The methods were modified to include photo neutron concentration. PARET 7.3 was used to simulate the transients behaviour of Ghana Research Reactor-1. The PARET code was capable of simulating the transients for 2.1 mk and 4 mk insertions of reactivity with peak powers of 49.87 kW and 92.34 kW, respectively. PARET code however failed to simulate 6.71 mk of reactivity which was predicted by Akaho et al through TEMPFED. (au)

  9. STAR: The Secure Transportable Autonomous Reactor System - Encapsulated Fission Heat Source

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2003-01-01

    OAK-B135 The Encapsulated Nuclear Heat Source (ENHS) is a novel 125 MWth fast spectrum reactor concept that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor. It uses Pb-Bi or other liquid-metal coolant and is intended to be factory manufactured in large numbers to be economically competitive. It is anticipated to be most useful to developing countries. The US team studying the feasibility of the ENHS reactor concept consisted of the University of California, Berkeley, Argonne National Laboratory (ANL), Lawrence Livermore National Laboratory (LLNL) and Westinghouse. Collaborating with the US team were three Korean organizations: Korean Atomic Energy Research Institute (KAERI), Korean Advanced Institute for Science and Technology (KAIST) and the University of Seoul, as well as the Central Research Institute of the Electrical Power Industry (CRIEPI) of Japan. Unique features of the ENHS include at least 20 years of operation without refueling; no fuel handling in the host country; no pumps and valves; excess reactivity does not exceed 1$; fully passive removal of the decay heat; very small probability of core damaging accidents; autonomous operation and capability of load-following over a wide range; very long plant life. In addition it offers a close match between demand and supply, large tolerance to human errors, is likely to get public acceptance via demonstration of superb safety, lack of need for offsite response, and very good proliferation resistance. The ENHS reactor is designed to meet the requirements of Generation IV reactors including sustainable energy supply, low waste, high level of proliferation resistance, high level of safety and reliability, acceptable risk to capital and, hopefully, also competitive busbar cost of electricity

  10. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

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

  11. Development of an educational nuclear research reactor simulator

    International Nuclear Information System (INIS)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim; Ashoub, Nagieb

    2014-01-01

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  12. Ground testing of an SP-100 prototypic reactor

    International Nuclear Information System (INIS)

    Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

    1988-01-01

    SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

  13. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Reactor Physics Dept.

    2014-12-15

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  14. 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)

  15. An aqueous lithium salt blanket option for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steiner, D.; Varsamis, G. (Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering and Engineering Physics); Deutsch, L.; Rathke, J. (Grumman Corp., Bethpage, NY (USA). Advanced Energy Systems); Gierszewski, P. (Canadian Fusion Fuels Technology Project (CFFTP), Mississauga, ON (Canada))

    1989-04-01

    An aqueous lithium salt blanket (ALSB) concept is proposed which could be the basis for either a power reactor blanket or a test module in an engineering test reactor. The design is based on an austenitic stainless steel structure, a beryllium multiplier, and a salt breeder concentration of about 32 g LiNO/sub 3/ per 100 cm/sup 3/ of H/sub 2/O. To limit tritium release rates, the salt breeder solution is separated from the water coolant circuit. The overall tritium system cost for a 2400 MW (fusion power) reactor is estimated to be 180 million Dollar US87 installed. (orig.).

  16. Neutron activation analysis of essential elements in Multani mitti clay using miniature neutron source reactor

    International Nuclear Information System (INIS)

    Waheed, S.; Rahman, S.; Faiz, Y.; Siddique, N.

    2012-01-01

    Multani mitti clay was studied for 19 essential and other elements. Four different radio-assay schemes were adopted for instrumental neutron activation analysis (INAA) using miniature neutron source reactor. The estimated weekly intakes of Cr and Fe are high for men, women, pregnant and lactating women and children while intake of Co is higher in adult categories and Mn by pregnant women. Comparison of MM clay with other type of clays shows that it is a good source of essential elements. - Highlights: ► Multani mitti clay has been studied for 19 essential elements for human adequacy and safety using INAA and AAS. ► Weekly intakes for different consumer categories have been calculated and compared with DRIs. ► Comparison of MM with other type of clays depict that MM clay is a good source of essential elements.

  17. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

    International Nuclear Information System (INIS)

    Chambadal, P.; Pascal, M.

    1955-01-01

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [fr

  18. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  19. The integral fast reactor - an overview

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Hannum, W.H.

    1997-01-01

    The Integral Fast Reactor (IFR) is a system that consists of a fast-spectrum nuclear reactor that uses metallic fuel and liquid-metal (sodium) cooling, coupled with technology for high-temperature electrochemical recycling, and with processes for preparing wastes for disposition. The concept is based on decades of experience with fast reactors, adapted to priorities that have evolved markedly from those of the early days of nuclear power. It has four essential, distinguishing features: efficient use of natural resources, inherent safety characteristics, reduced burdens of nuclear waste, and unique proliferation resistance. These fundamental characteristics offer benefits in economics and environmental protection. The fuel cycle never involves separated plutonium, immediately simplifying the safeguarding task. Initiated in 1984 in response to proliferation concerns identified in the International Nuclear Fuel Cycle Evaluation (INFCE, 1980), the project has made substantial technical progress, with new potential applications coming to light as nuclear weapons stockpiles are reduced and concerns about waste disposal increase. A breakthrough technology, the IFR has the characteristics necessary for the next nuclear age. (author)

  20. An overview of future sustainable nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Poullikkas, Andreas [Electricity Authority of Cyprus, P.O. Box 24506, 1399 Nicosia (Cyprus)

    2013-07-01

    In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will tend to have closed

  1. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Science.gov (United States)

    Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

    2015-12-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

  2. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    International Nuclear Information System (INIS)

    Hamann, S.; Röpcke, J.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.

    2015-01-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH 4 , C 2 H 2 , HCN, and NH 3 ). With the help of OES, the rotational temperature of the screen plasma could be determined

  3. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

    2015-12-15

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

  4. Physical models and primary design of reactor based slow positron source at CMRR

    Science.gov (United States)

    Wang, Guanbo; Li, Rundong; Qian, Dazhi; Yang, Xin

    2018-07-01

    Slow positron facilities are widely used in material science. A high intensity slow positron source is now at the design stage based on the China Mianyang Research Reactor (CMRR). This paper describes the physical models and our primary design. We use different computer programs or mathematical formula to simulate different physical process, and validate them by proper experiments. Considering the feasibility, we propose a primary design, containing a cadmium shield, a honeycomb arranged W tubes assembly, electrical lenses, and a solenoid. It is planned to be vertically inserted in the Si-doping channel. And the beam intensity is expected to be 5 ×109

  5. Moderators for the design of a cold neutron source for the RA 3 reactor

    International Nuclear Information System (INIS)

    Cantargi, F; Sbaffoni, M; Granada, R

    2004-01-01

    The cold neutron production of hydrogenous materials was studied, taking into account their radiation resistance, for the conceptual design of a cold neutron source for the RA-3 reactor.Low spontaneous release of chemical energy was found in mesitylene.Libraries for hidrogen in mesitylene were generated using the NJOY nuclear processing system and the resulting cross sections were compared with experimental data.Good agreement between measurements and calculations was found in those cases where data are available.New calculations using the RA-3 geometry and these validated libraries will be performed [es

  6. A safe private nuclear tool-the miniature neutron source reactor

    International Nuclear Information System (INIS)

    Zhou Yongmao

    1987-01-01

    The prototype miniature neutron source reactor (MNSR) designed by China Institute of Atomic Energy has been operated successfully for more than 3 years and the practical experience enriches the original design idea. The commercial MNSR is under study design and develop in following aspects: 1. Prolonging the control rod cycle duration and core burn-up life; 2. Increasing the neutron flux per unit power. Obviously, the MNSR will show more advantages in extending application area and in providing the core using low envichment fuel. (Liu)

  7. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  8. An expert system for pressurized water reactor load maneuvers

    International Nuclear Information System (INIS)

    Chaung Lin; Jungping Chen; Yihjiunn Lin; Lianshin Lin

    1993-01-01

    Restartup after reactor shutdown and load-follow operations are the important tasks in operating pressurized water reactors. Generally, the most efficient method is to apply constant axial offset control (CAOC) strategy during load maneuvers. An expert system using CAOC strategy, fuzzy reasoning, a two-node core model, and operational constraints has been developed. The computation time is so short that this system, which leads to an approximate closed-loop control, could be useful for on-site calculation

  9. An optimization method for parameters in reactor nuclear physics

    International Nuclear Information System (INIS)

    Jachic, J.

    1982-01-01

    An optimization method for two basic problems of Reactor Physics was developed. The first is the optimization of a plutonium critical mass and the bruding ratio for fast reactors in function of the radial enrichment distribution of the fuel used as control parameter. The second is the maximization of the generation and the plutonium burnup by an optimization of power temporal distribution. (E.G.) [pt

  10. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    International Nuclear Information System (INIS)

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-01-01

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

  11. Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

    International Nuclear Information System (INIS)

    Gehin, J.C.; Worley, B.A.; Renier, J.P.; Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M.

    1995-08-01

    This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates

  12. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    Science.gov (United States)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  13. Ceramic membrane microfilter as an immobilized enzyme reactor.

    Science.gov (United States)

    Harrington, T J; Gainer, J L; Kirwan, D J

    1992-10-01

    This study investigated the use of a ceramic microfilter as an immobilized enzyme reactor. In this type of reactor, the substrate solution permeates the ceramic membrane and reacts with an enzyme that has been immobilized within its porous interior. The objective of this study was to examine the effect of permeation rate on the observed kinetic parameters for the immobilized enzyme in order to assess possible mass transfer influences or shear effects. Kinetic parameters were found to be independent of flow rate for immobilized penicillinase and lactate dehydrogenase. Therefore, neither mass transfer nor shear effects were observed for enzymes immobilized within the ceramic membrane. Both the residence time and the conversion in the microfilter reactor could be controlled simply by regulating the transmembrane pressure drop. This study suggests that a ceramic microfilter reactor can be a desirable alternative to a packed bed of porous particles, especially when an immobilized enzyme has high activity and a low Michaelis constant.

  14. An innovative nuclear reactor as a solution to global warming

    International Nuclear Information System (INIS)

    Silva, Robson Silva da; Sefidvash, Farhang

    2007-01-01

    The problem of global warming is no longer a philosophical discussion, but it is a fact seriously threatening the future of humanity. In this paper a practical solution to the problem of global warming resulting from the fossil fuelled energy suppliers is presented. The energy conservation and alternative forms of energy such as solar, wind, and bio even though having important roles, do not satisfy the energy demand generated by an increasing world population that desires to increase its standard of living. The fission process in the nuclear reactors does not produce greenhouse gases that cause global warming. The new paradigm in nuclear energy is the future innovative reactors that meet the new standards set by the INPRO Program of the IAEA. One such a reactor is presented in this paper, namely the Fixed Bed Nuclear Reactor (FBNR) that is supported by the International Atomic Energy (IAEA) in its program of Small Reactors Without On-Site Refuelling (SRWOSR), being one of the four water cooled reactors in this program. The other three reactor concepts are PFPWR50 of Japan, BWRPB of Russia and AFPR-100 of USA. It is shown that the nuclear energy of the future is totally different than what is today in respect to safety, economics, environmental impact and proliferation. In this manner, the public perception of nuclear energy will change and its acceptability is promoted. (author)

  15. Synthetic fuel production via carbon neutral cycles with high temperature nuclear reactors as a power source

    Energy Technology Data Exchange (ETDEWEB)

    Konarek, E.; Coulas, B.; Sarvinis, J. [Hatch Ltd., Mississauga, Ontario (Canada)

    2016-06-15

    This paper analyzes a number of carbon neutral cycles, which could be used to produce synthetic hydrocarbon fuels. Synthetic hydrocarbons are produced via the synthesis of Carbon Monoxide and Hydrogen. The . cycles considered will either utilize Gasification processes, or carbon capture as a source of feed material. In addition the cycles will be coupled to a small modular Nuclear Reactor (SMR) as a power and heat source. The goal of this analysis is to reduce or eliminate the need to transport diesel and other fossil fuels to remote regions and to provide a carbon neutral, locally produced hydrocarbon fuel for remote communities. The technical advantages as well as the economic case are discussed for each of the cycles presented. (author)

  16. Synthetic fuel production via carbon neutral cycles with high temperature nuclear reactors as a power source

    International Nuclear Information System (INIS)

    Konarek, E.; Coulas, B.; Sarvinis, J.

    2016-01-01

    This paper analyzes a number of carbon neutral cycles, which could be used to produce synthetic hydrocarbon fuels. Synthetic hydrocarbons are produced via the synthesis of Carbon Monoxide and Hydrogen. The . cycles considered will either utilize Gasification processes, or carbon capture as a source of feed material. In addition the cycles will be coupled to a small modular Nuclear Reactor (SMR) as a power and heat source. The goal of this analysis is to reduce or eliminate the need to transport diesel and other fossil fuels to remote regions and to provide a carbon neutral, locally produced hydrocarbon fuel for remote communities. The technical advantages as well as the economic case are discussed for each of the cycles presented. (author)

  17. Modification to ORIGEN2 for generating N Reactor source terms. Volume 1

    International Nuclear Information System (INIS)

    Schwarz, R.A.

    1997-04-01

    This report discusses work that has been done to upgrade the ORIGEN2 code cross sections to be compatible with the WIMS computer code data. Because of the changes in the ORIGEN2 calculations. Details on changes made to the ORIGEN2 computer code and the Radnuc code will be discussed along with additional work that should be done in the future to upgrade both ORIGEN2 and Radnuc. A detailed historical description of how source terms have been generated for N Reactor fuel stored in the K Basins has been generated. The neutron source discussed in this description was generated by the WIMS computer code (Gubbins et al. 1982) because of known shortcomings in the ORIGEN2 (Croff 1980) cross sections. Another document includes a discussion of the ORIGEN2 cross sections

  18. Study of the External Neutron Source Effect on TRU Burning in a Sub-critical Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zafar, Zafar Iqbal; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    One of the drawback points of nuclear power is the production of highly radioactive and long lasting waste isotopes during power production. Therefore, most important design requirement of future nuclear option should have a potential to burn selectively long-lived fission products (LLFP) and long-lived minor actinides (LLMA). However, there is no way to burn them selectively in the reactor core. Practical method of waste transmutation should rely on selective separation of them from spent nuclear fuel of power plants. Under the proliferation concern, direct separation of trans-uranic isotopes (TRU) from pyro-reprocessing plant became a feasible option in our country. Even though social political agreement is not matured as well as technical feasibility, current study is done based on basic assumptions; TRU and LLFP is separated from spent fuel of nuclear power plants. The remaining neutrons (among the external 3%) very few in number (less than 1% in any case) being very energetic (above three MeV or so) do cause much more fissions per neutron than their counterparts but, because of their overall low population they do not have any significant and decisive influence in the overall reactor performance. Currently, entire study is limited to the source neutron energy of 20 MeV only. In future, it is expected to get reasonably plausible fixed source dependent difference in the TRU burning by using tabulated data for the neutrons of higher energy (up to 250 MeV at least). Secondly, a clearer picture is expected if the TRU loading was increased from the current value of 133 kg to few metric tons, as is the case in most of the existing reactors.

  19. Pressing device for producing compacts from source material in powder form in particular pulverized nuclear reactor fuel

    International Nuclear Information System (INIS)

    Heller, G.; Adelmann, M.; Konigs, W.; Wendorf, W.

    1984-01-01

    Pressing device for producing compacts from source material in powder form, in particular pulverized nuclear reactor fuel having a die-plate contained in platen and a bore associated with a ram, for receiving source material powder, a filling shoe, and a reservoir for powder connected by a hose to the filling shoe. The device is characterized by a passing wheel in the filling shoe as filling aid means; a tube containing a feedscrew disposed between the reservoir and hose as metering means; the reservoir having a bottom part with a can type place-on part with an opening eccentric to the axis; a coupling part and a cover part are placed on the open part of the can, these parts are also provided with a passageway to the feedscrew eccentric to the longitudinal axis

  20. Estimation of the sub-criticality of the sodium-cooled fast reactor Monju using the modified neutron source multiplication method

    International Nuclear Information System (INIS)

    Truchet, G.; Van Rooijen, W. F. G.; Shimazu, Y.; Yamaguchi, K.

    2012-01-01

    The Modified Neutron Source Method (MNSM) is applied to the Monju reactor. This static method to estimate sub-criticality has already given good results on commercial Pressurized Water Reactors. The MNSM consists both in the extraction of the fundamental mode seen by a detector to avoid the effect of higher modes near sources, and the correction of flux distortion effects due to control rod movement. Among Monju's particularities that have a big influence on MNSM factors are: the presence of two californium sources and the position of the detector which is located far from the core outside of the reactor vessel. The importance of spontaneous fission and (α, n) reactions which have increased during the shutdown period of 15 years will also be discussed. The relative position of detectors and sources deeply affect the correction factors in some regions. In order to evaluate the detector count rate, an analytical propagation has been conducted from the reactor vessel. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). (authors)

  1. Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

    International Nuclear Information System (INIS)

    Khattab, K.; Omar, H.; Ghazi, N.

    2009-01-01

    A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

  2. Measurement of Relative Biological Effectiveness (RBE) for the Radiation Beam from Neutron Source Reactor YAYOI -Comparisons with Cyclotron Neutron and 60Co Gamma Ray-

    OpenAIRE

    HIROAKI, WAKABAYASHI; SHOZO, SUZUKI; AKIRA, ITO; Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo; Institute of Medical Science, the University of Tokyo; Institute of Medical Science, the University of Tokyo

    1983-01-01

    Radiation biology and/or therapy research and development for a research reactor beam need specific RBEs of neutrons as well as of specific reactions. RBEs for reactor beams measured in situ condition are interesting because actual radiation effects on each biological system are different depending on detailed conditions of irradiation. A small powered research reactor (Fast Neutron Source Reactor: YAYOI) was examined here as a neutron beam source for obtaining survival curves in a manner usu...

  3. Trip setpoint analysis for the reactor protection system of an advanced integral reactor

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Kim, Soo Hyung; Chung, Young Jong; Zee, Sung Quun

    2007-01-01

    The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria

  4. Hydrodynamic study of an internal airlift reactor for microalgae culture.

    Science.gov (United States)

    Rengel, Ana; Zoughaib, Assaad; Dron, Dominique; Clodic, Denis

    2012-01-01

    Internal airlift reactors are closed systems considered today for microalgae cultivation. Several works have studied their hydrodynamics but based on important solid concentrations, not with biomass concentrations usually found in microalgae cultures. In this study, an internal airlift reactor has been built and tested in order to clarify the hydrodynamics of this system, based on microalgae typical concentrations. A model is proposed taking into account the variation of air bubble velocity according to volumetric air flow rate injected into the system. A relationship between riser and downcomer gas holdups is established, which varied slightly with solids concentrations. The repartition of solids along the reactor resulted to be homogenous for the range of concentrations and volumetric air flow rate studied here. Liquid velocities increase with volumetric air flow rate, and they vary slightly when solids are added to the system. Finally, liquid circulation time found in each section of the reactor is in concordance with those employed in microalgae culture.

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

    International Nuclear Information System (INIS)

    Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

    1991-01-01

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

  6. Antimatter as an Energy Source

    International Nuclear Information System (INIS)

    Jackson, Gerald P.

    2009-01-01

    Antiprotons and positrons are constantly generated in space, and periodically manufactured by humans here on Earth. Harvesting of these particles in space and forming stable antimatter atoms and molecules would create a significant energy source for power and propulsion. Though dedicated fabrication of these particles on Earth consumes much more energy than could be liberated upon annihilation, manufactured antimatter represents a high-density energy storage mechanism well suited for spacecraft power and propulsion. In this paper the creation, storage, and utilization of antimatter is introduced. Specific examples of electrical energy generation and deep-space propulsion based on antimatter are also reviewed.

  7. Investigation of a superthermal ultracold neutron source based on a solid deuterium converter for the TRIGA Mainz reactor

    International Nuclear Information System (INIS)

    Lauer, Thorsten

    2010-01-01

    Research in fundamental physics with the free neutron is one of the key tools for testing the Standard Model at low energies. Most prominent goals in this field are the search for a neutron electric dipole moment (EDM) and the measurement of the neutron lifetime. Significant improvements of the experimental performance using ultracold neutrons (UCN) require reduction of both systematic and statistical errors.The development and construction of new UCN sources based on the superthermal concept is therefore an important step for the success of future fundamental physics with ultracold neutrons. Significant enhancement of today available UCN densities strongly correlates with an efficient use of an UCN converter material. The UCN converter here is to be understood as a medium which reduces the velocity of cold neutrons (CN, velocity of about 600 m/s) to the velocity of UCN (velocity of about 6 m/s).Several big research centers around the world are presently planning or constructing new superthermal UCN sources, which are mainly based on the use of either solid deuterium or superfluid helium as UCN converter.Thanks to the idea of Yu.Pokotilovsky, there exists the opportunity to build competitive UCN sources also at small research reactors of the TRIGA type. Of course these smaller facilities don't promise high UCN densities of several 1000 UCN/cm 3 , but they are able to provide densities around 100 UCN/cm 3 for experiments.In the context of this thesis, it was possible to demonstrate succesfully the feasibility of a superthermal UCN source at the tangential beamport C of the research reactor TRIGA Mainz. Based on a prototype for the future UCN source at the Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII) in Munich, which was planned and built in collaboration with the Technical University of Munich, further investigations and improvements were done and are presented in this thesis. In parallel, a second UCN source for the radial beamport D was designed and

  8. Investigation of a superthermal ultracold neutron source based on a solid deuterium converter for the TRIGA Mainz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lauer, Thorsten

    2010-12-22

    Research in fundamental physics with the free neutron is one of the key tools for testing the Standard Model at low energies. Most prominent goals in this field are the search for a neutron electric dipole moment (EDM) and the measurement of the neutron lifetime. Significant improvements of the experimental performance using ultracold neutrons (UCN) require reduction of both systematic and statistical errors.The development and construction of new UCN sources based on the superthermal concept is therefore an important step for the success of future fundamental physics with ultracold neutrons. Significant enhancement of today available UCN densities strongly correlates with an efficient use of an UCN converter material. The UCN converter here is to be understood as a medium which reduces the velocity of cold neutrons (CN, velocity of about 600 m/s) to the velocity of UCN (velocity of about 6 m/s).Several big research centers around the world are presently planning or constructing new superthermal UCN sources, which are mainly based on the use of either solid deuterium or superfluid helium as UCN converter.Thanks to the idea of Yu.Pokotilovsky, there exists the opportunity to build competitive UCN sources also at small research reactors of the TRIGA type. Of course these smaller facilities don't promise high UCN densities of several 1000 UCN/cm{sup 3}, but they are able to provide densities around 100 UCN/cm{sup 3} for experiments.In the context of this thesis, it was possible to demonstrate succesfully the feasibility of a superthermal UCN source at the tangential beamport C of the research reactor TRIGA Mainz. Based on a prototype for the future UCN source at the Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII) in Munich, which was planned and built in collaboration with the Technical University of Munich, further investigations and improvements were done and are presented in this thesis. In parallel, a second UCN source for the radial beamport D was

  9. Efficiency of an LBE spallation target in an accelerator-driven molten salt subcritical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bak, Sang-In [Sungkyunkwan University, Suwon (Korea, Republic of); Hong, Seung-Woo [Sungkyunkwan University, Suwon (Korea, Republic of); Kadi, Yacine [CERN, Geneva (Switzerland)

    2016-10-15

    An Accelerator-Driven System (ADS) combined with a subcritical Molten Salt Reactor (MSR) is a type of hybrid reactor originally designed to breed uranium from thorium or to incinerate long-lived minor actinides in nuclear wastes. In an MSR, the salt material is used not only as a nuclear fuel but also as a primary coolant. In addition, this material is used as a target for inducing spallation neutrons in most AD-MSR concepts. A high energy proton beam impinges on a heavy metal target to induce spallation reactions and produces neutrons. Accordingly, a reliable proton accelerator is needed to feed the source neutrons. As ADSs have been criticized for requiring high power accelerators, minimization of beam power is an important aspect of ADS design. A primary concern associated with ADS development is stable high-power accelerators. We therefore studied the neutron source efficiencies of an AD-MSR involving chloride fuels by including a Pb-Bi eutectic (LBE) spallation target. The proton source efficiency and the accelerator beam power required have been studied for an AD-MSR. Adoption of an LBE spallation target induces an increase in proton source efficiencies in comparison to the case without a spallation target. Thus the presence of an efficient spallation target is useful in the reduction of the beam power of an accelerator. Almost 33 % of the beam power can be reduced in comparison to the case without the target for NaCl-Th/{sup 233}U fuel, and about 16 % for NaCl-U/TRU fuel. The beam power amplifications increase by 1.5 times for NaCl-Th/{sup 233}U and 1.2 times for NaCl-U/TRU in comparison with the no target AD-MSR.

  10. Small-angle scattering at a pulsed neutron source: comparison with a steady-state reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borso, C S; Carpenter, J M; Williamson, F S; Holmblad, G L; Mueller, M H; Faber, J Jr; Epperson, J E; Danyluk, S S [Argonne National Lab., IL (USA)

    1982-08-01

    A time-of-flight small-angle diffractometer employing seven tapered collimator elements and a two-dimensional gas proportional counter was successfully utilized to collect small-angle scattering data from a solution sample of the lipid salt cetylpyridinium chloride, C/sub 21/H/sub 38/N/sup +/.Cl/sup -/, at the Argonne National Laboratory prototype pulsed spallation neutron source, ZING-P'. Comparison of the small-angle scattering observed from the same compound at the University of Missouri Research Reactor corroborated the ZING-P' results. The results are used to compare the neutron flux available from the ZING-P' source relative to the well characterized University of Missouri source. Calculations based on experimentally determined parameters indicated the time-averaged rate of detected neutrons at the ZING-P' pulsed spallation source to have been at least 33% higher than the steady-state count rate from the same sample. Differences between time-of-flight techniques and conventional steady-state techniques are discussed.

  11. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  12. Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The order is enacted under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. Any person who engages in refining business shall get designation for each works or place of enterprise. The application shall be filed through the director of International Trade and Industry Office in charge of the location of the works or the enterprise with a business program and other specified documents attached. Any person who undertakes processing business shall get permission for each works or place of enterprise. The application shall be submitted with a business program and other documents defined by the Ordinance of the Prime Minister's Office. Any person who sets up reactor shall get permission for each works or place of enterprise. The application shall be presented with a financial project and other documents stipulated by the ordinance. Fast breeding reactor, heavy-water moderated boiling water reactor and light-water moderated pressurized water reactor are designated as reactor in the phase of research and development. Each foreign nuclear ship equipped with reactor which enters into Japanese waters shall get permission of the Minister of Transport. The application shall be presented with the papers explaining safety of reactor facilities and other documents provided by the ordinance of the ministry concerned. (Okada, K.)

  13. Processing test of an upgraded mechanical design for PERMCAT reactor

    International Nuclear Information System (INIS)

    Borgognoni, Fabio; Demange, David; Doerr, Lothar; Tosti, Silvano; Welte, Stefan

    2010-01-01

    The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H 2 O and D 2 .

  14. An overview of the RECH-1 reactor conversion

    International Nuclear Information System (INIS)

    Klein, J.; Medel, J.; Daie, J.; Torres, H.

    2000-01-01

    The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm 3 . This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U 3 Si 2 -Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm 3 . A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)

  15. Neutronics-design modification of the Syrian miniature neutron source reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.

    2001-01-01

    An investigation to improve performance of the MNSR is presented. Modification of the available excess reactivity of the core through the addition of 2 mk reactivity is suggested. New operating procedure of the reactor is devised. Measured and calculated effect of temperature reactivity coefficient was evaluated. Good agreement between measured and calculated reactivity effects was obtained. Safety margin for the modified case has been slightly enhanced. (Author)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  17. Prospects for inertial fusion as an energy source

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1989-01-01

    Progress in the Inertial Confinement Fusion (ICF) Program has been very rapid in the last few years. Target physics experiments with laboratory lasers and in underground nuclear tests have shown that the drive conditions necessary to achieve high gain can be achieved in the laboratory with a pulse-shaped driver of about 10 MJ. Requirements and designs for a Laboratory Microfusion Facility (LMF) have been formulated. Research on driver technology necessary for an ICF reactor is making progress. Prospects for ICF as an energy source are very promising. 11 refs., 5 figs

  18. Lactose hydrolysis in an enzymatic membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mertens, B; Huyghebaert, A

    1987-10-01

    The enzymatic hydrolysis of lactose in whey permeate with subsequent recuperation of Saccharomyces lactis lactase by means of ultrafiltration was investigated. In whey permeate, S. lactis lactase shows maximal activity at pH 6.5; the optimal temperature was found to be 45/sup 0/C and is limited by strong thermal inactivation beyond this temperature. High activity combined with acceptable thermal inactivation (< 10% after 5 h incubation) was established at 30/sup 0/C. S. lactis lactase also displays considerable activity at low temperature (5/sup 0/C). Enzyme stability is reduced drastically by demineralisation: addition of low concentrations of manganese ions (10/sup -3/ M) considerably enhances stability. Using a DDS Lab-Unit 35 fitted with GR61PP polysulphon membranes (cut-off: 20.000), pilot scale experiments were carried out (pH 6.5; 30/sup 0/C) in which whey permeate was hydrolyzed to a degree of hydrolysis of 82% minimum. Enzyme recuperation amounted to 96.5% per batch, all enzyme activity loss being due to thermal inactivation. Microbiological examination of the enzymatic membrane reactor showed that growth of mcicroorganisms can largely be suppressed by working at lower temperature (5/sup 0/C). Eventually, 50 ppm H/sub 2/O/sub 2/ or sterile filtration will adequately solve microbiological problems without affecting enzyme activity.

  19. Multi-purpose nuclear heat source for advanced gas-cooled reactor plants

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1993-01-01

    Nuclear power has the potential to be the ultimate green technology in that it could eliminate the need for burning fossil fuels with their polluting combustion products and greenhouse gases. This view is shared by many technologists, but it may be a generation before the public becomes convinced, and that will involve overcoming many safety, institutional, financial, and technical impediments. This paper addresses only the latter topic; a major theme being that for nuclear power to truly be a green technology and significantly benefit society, it must meet the needs of the full energy spectrum. Specifically, it must satisfy energy needs beyond just the electricity generating sector by today's nuclear plants. By virtue of its high temperature capability, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is the only type of reactor that has the potential to meet the wide range of energy needs that will emerge in the future. This paper discusses the nuclear heat source that gives the MHTGR multi-purpose capability, which is recognized today, but will not be implemented until early in the next century

  20. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  1. An improved water cooled nuclear reactor and pressuriser assembly

    International Nuclear Information System (INIS)

    Gardner, F.J.; Strong, R.

    1991-01-01

    A water cooled nuclear reactor is described which comprises a reactor core, a primary water coolant circuit and a pressuriser arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper and a lower chamber by a casing. The reactor core and primary water coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber. A plurality of spray pipes interconnect a steam space of the pressuriser with the downcomer of the primary water coolant circuit below a heat exchanger. A plurality of surge ports interconnect a water space of the pressuriser with the primary water coolant circuit. The surge ports have hydraulic diodes so that there is a low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and high flow resistance in the opposite direction. The spray pipes provide a desuperheating spray of cooled water into the pressuriser during positive volume surges of the primary water coolant. The pressuriser arrangement may also be applied to integral water cooled reactors with separate pressurisers and to dispersed pressurised water reactors. The surge ports also allow water to flow by gravity to the core in an emergency. (author)

  2. Project of SVBR-75/100 reactor plant with improved safety for nuclear sources of small and medium power

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Stepanov, V. S.; Klimov, N. N.; Dedul, A. V.; Bolvanchikov, S. N.; Zrodnikov, A. V.; Tolhinsky, G. I.; Komlev, O. G.

    2004-01-01

    As a result of the joint work performed recently by FSUE OKB Gidropress, SNC RF-IPPE and other organizations the technical feasibility is shown for creation and usage in nuclear power engineering of the unified reactor plant (RP) SVBR-75/100 with fast neutron reactor core and lead-bismuth coolant (LBC) in the primary circuit. Technical design of SVBR-75/100 reactor plant is based on the following: 50-year operation experience in development and operation of RP with LBC for nuclear submarines; experience in development and operation of fast reactor with sodium coolant; experience in optimization of LBC technology at nuclear submarines and ground-based test benches; conceptual design of SVBR-75 reactor plant (for renovation of Units 2, 3 and 4 of Novovoronezh NPP). Technical solutions laid down in the basis of SVBR-75/100 reactor plant design are oriented towards the industrial basis, structural materials existing in Russia, as well as the unique LBC technology with experimental and practical support. The concept of SVBR-75/100 reactor plant safety assurance is based on the following provisions: maximum usage of inherent safety supported by physical features of fast neutron reactor, chemically inert LBC in the primary circuit, integral layout and special design solutions; maximum possible combination of normal operation and safety functions in RP systems. Small power of SVBR-75/100 RP makes it possible to manufacture the complete set of RP main equipment at the factory and delivery it to NPP site as-finished practically using any transport including railway. Possible fields of application of SVBR-75/100 reactor plants: modular NPPs of different power; renovation of NPPs with light water reactors exhausted their service life; independent nuclear power sources for different applications (ground-based nuclear water-desalinating plants, etc. )(author)

  3. Study of cold and hot sources in a research reactor. (Physics, specifications, operation, utilization)

    International Nuclear Information System (INIS)

    Safieh, J.

    1982-10-01

    A brief description of the reactor, sources and experimental channels (ORPHEE being taken as example) is first given. The first part deals with the hot neutron source, mainly made of a graphite block to be carried at a temperature of 1500 0 K by nuclear heating. The present study focused on the determination, with the code MERCURE IV, of heat sources generated in the graphite block. From these results the spatial distribution of temperatures have been calculated with two different methods. Mechanical and thermal stresses have been calculated for the hot points. Then, the outlet neutron spectra is determined by means of the code APOLLO. Finally, the operation of the device is presented and the risks and the safety measures are given. The second part deals with cold neutron sources comprising mainly a cold moderator (liquid hydrogen 20.4 0 K). The helium coolant circuit liquefies the hydrogen by means of heat exchange in a condenser. Cold neutron yields calculations are developed by means of the code THERMOS in the plane and cyclindrical geometries. Heat sources generated by nuclear radiations are calculated. A detailed description of the device and its coolant circuit is given, and a risk analysis is finally presented. The third part deals with the part of thermal cold and hot neutrons in the study of matter and its dynamics. Technical means needed to obtain a monochromatic beam, for diffraction experiments, are recalled emphasizing on the interest of these neutrons with regard to X radiation. Then, one deals with cold neutron guides. Finally, the efficiency of two neutron guides is calculated. 78 refs [fr

  4. Monte Carlo neutronics analysis of the ANS reactor three-element core design

    International Nuclear Information System (INIS)

    Wemple, C.A.

    1995-01-01

    The advanced neutron source (ANS) is a world-class research reactor and experimental center for neutron research, currently being designed at the Oak Ridge National Laboratory (ORNL). The reactor consists of a 330-MW(fission) highly enriched uranium core, which is cooled, moderated, and reflected with heavy water. It was designed to be the preeminent ultrahigh neutron flux reactor in the world, with facilities for research programs in biology, materials science, chemistry, fundamental and nuclear physics, and analytical chemistry. Irradiation facilities are provided for a variety of isotope production capabilities, as well as materials irradiation. This paper summarizes the neutronics efforts at the Idaho National Engineering Laboratory in support of the development and analysis of the three-element core for the advanced conceptual design phase

  5. Evaluation of marine algae as a source of biogas in a two-stage anaerobic reactor system

    International Nuclear Information System (INIS)

    Vergara-Fernandez, Alberto; Vargas, Gisela; Alarcon, Nelson; Velasco, Antonio

    2008-01-01

    The marine algae are considered an important biomass source; however, their utilization as energy source is still low around the world. The technical feasibility of marine algae utilization as a source of renewable energy was studied to laboratory scale. The anaerobic digestion of Macrocystis pyrifera, Durvillea antarctica and their blend 1:1 (w/w) was evaluated in a two-phase anaerobic digestion system, which consisted of an anaerobic sequencing batch reactor (ASBR) and an upflow anaerobic filter (UAF). The results show that 70% of the total biogas produced in the system was generated in the UAF, and both algae species have similar biogas productions of 180.4(±1.5) mL g -1 dry algae d -1 , with a methane concentration around 65%. The same methane content was observed in biogas yield of algae blend; however, a lower biogas yield was obtained. In conclusion, either algae species or their blend can be utilized to produce methane gas in a two-phase digestion system

  6. Evaluation of marine algae as a source of biogas in a two-stage anaerobic reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Vergara-Fernandez, Alberto; Vargas, Gisela [Escuela de Ingenieria Ambiental, Facultad de Ingenieria, Universidad Catolica de Temuco, Manuel Montt 56, Casilla 15-D, Temuco (Chile); Alarcon, Nelson [Departamento de Ingenieria Quimica, Facultad de Ingenieria y Ciencias Geologicas, Universidad Catolica del Norte (Chile); Velasco, Antonio [Centro Nacional de Investigacion y Capacitacion Ambiental del Instituto Nacional de Ecologia (CENICA-INE), Av. San Rafael Atlixco 186, Col. Vicentina, Del. Iztapalapa, 09340, Mexico, DF (Mexico)

    2008-04-15

    The marine algae are considered an important biomass source; however, their utilization as energy source is still low around the world. The technical feasibility of marine algae utilization as a source of renewable energy was studied to laboratory scale. The anaerobic digestion of Macrocystis pyrifera, Durvillea antarctica and their blend 1:1 (w/w) was evaluated in a two-phase anaerobic digestion system, which consisted of an anaerobic sequencing batch reactor (ASBR) and an upflow anaerobic filter (UAF). The results show that 70% of the total biogas produced in the system was generated in the UAF, and both algae species have similar biogas productions of 180.4({+-}1.5) mL g{sup -1} dry algae d{sup -1}, with a methane concentration around 65%. The same methane content was observed in biogas yield of algae blend; however, a lower biogas yield was obtained. In conclusion, either algae species or their blend can be utilized to produce methane gas in a two-phase digestion system. (author)

  7. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  8. Experimental evaluation of an expert system for nuclear reactor operators

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1984-10-01

    The United States Nuclear Regulatory Commission (USNRC) is supporting a program for the experimental evaluation of an expert system for nuclear reactor operators. A prototype expert system, called the Response Tree System, has been developed and implemented at INEL. The Response Tree System is designed to assess the status of a reactor system following an accident and recommend corrective actions to reactor operators. The system is implemented using color graphic displays and is driven by a computer simulation of the reactor system. Control of the system is accomplished using a transparent touch panel. Controlled experiments are being conducted to measure performance differences between operators using the Response Tree System and those not using it to respond to simulated accident situations. This paper summarizes the methodology and results of the evaluation of the Response Tree System, including the quantitative results obtained in the experiments thus far. Design features of the Response Tree System are discussed, and general conclusions regarding the applicability of expert systems in reactor control rooms are presented

  9. Foundational development of an advanced nuclear reactor integrated safety code

    International Nuclear Information System (INIS)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-01-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  10. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  11. An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

    NARCIS (Netherlands)

    Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

    2009-01-01

    In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

  12. An intelligent safety system concept for future CANDU reactors

    International Nuclear Information System (INIS)

    Hinds, H.W.

    1980-01-01

    A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)

  13. An elevator for cobalt-60 source

    International Nuclear Information System (INIS)

    Tang Zaimin; Liang Donghu

    1990-07-01

    The elevator used for cobalt-60 source is a key device in the irradiation industry. It plays an important role in the safety and control of irradiation operation as well as the utilization rate of radiation source. From 1983 to 1986, Beijing Institute of Nuclear Engineering undertook designing of various size irradiation projects for different uses. Since then a kind of cobalt-60 source elevator suited for the irradiator of wet-source-storage has been chosen. It is reliable in the operation and complete in the function. An automatic control circuit brings the systems of cobalt-60 source elevator into an interlock system which ensures the irradiation operation safety. Besides introducing the structural features and performance of this elevator, the conditions of safety interlocking in raising or lowering the cobalt-60 source is also discussed. The discussion is from the safety viewpoint of operating an irradiator and irradiation technology

  14. An assessment of the radiological consequences of accidents in research reactors

    International Nuclear Information System (INIS)

    Ferreira, N.L.D.

    1992-01-01

    This work analyses the radiological consequences of accidents in two types of research reactors: a 5 MWt open pool reactor and a 50 MWt PWR reactor. Two siting cases have been considered: the reactor located near to a large population center and sited in a rural area. The influence of several factors such as source term, meteorological conditions and population distribution have been considered in the present analysis. (author)

  15. Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

    Science.gov (United States)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion

  16. Effects of neutron source ratio on nuclear characteristics of D-D fusion reactor blankets and shields

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Nakao, Yasuyuki; Ohta, Masao

    1978-01-01

    An examination is made of the dependence shown by the nuclear characteristics of the blanket and shield of D-D fusion reactors on S sub( d d)/S sub( d t), the ratio between the 2.45 MeV neutrons resulting from the D-D reaction and those of 14.06 MeV from the D-T reaction. Also, an estimate is presented of this neutron source ratio S sub( d d)/S sub( d t) for the case of D-D reactors, taken as an example. It is shown that an increase of S sub( d d)/S sub( d t) reduces the amount of nuclear heating per unit source neutron, while at the same time improving the shielding characteristics. This is accountable to lowering of the energy and penetrability of incident neutrons into the blanket brought about by the increase of S sub( d d)/S sub( d t). The value of S sub( d d)/S sub( d t) in a steady state D-D fusioning plasma core is estimated to be 1.46 -- 1.72 for an ion temperature ranging from 60 -- 180 keV. The reductions obtained on H sub( t)sup( b) (total heating in the blanket), H sub( t)sup( m g)/H sub( t)sup( b) (shielding indicator = ratio between total heating in superconducting magnet and that in the blanket) and phi sup( m g)/phi sup( w) (ratio of fast neutron fluxes between that at the magnet inner surface and that at the first wall inner surface) brought about by increasing S sub( d d)/S sub( d t) from unity to the value cited above do not differ to any appreciable extent, whichever is adopted among the design models considered here, the differences being at most about 10, 15 and 25%, respectively, for these three parameters. These results would broaden the validity of the conclusion derived in the previous paper for the case of S sub( d d)/S sub( d t) = 1.0, that the blanket-shield concept would appear to be the most suitable for D-D fusion reactors. (author)

  17. PX–An Innovative Safety Concept for an Unmanned Reactor

    Directory of Open Access Journals (Sweden)

    Sung-Jae Yi

    2016-02-01

    Full Text Available An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

  18. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  19. Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment

    KAUST Repository

    Zhu, Xiuping

    2013-05-01

    Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H2O2 generation and Fe2+ release for the Fenton reaction. In tests using phenol, 75±2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants. © 2013 Elsevier B.V.

  20. Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment

    International Nuclear Information System (INIS)

    Zhu, Xiuping; Logan, Bruce E.

    2013-01-01

    Highlights: ► A new type of electro-Fenton system was developed for wastewater treatment. ► Degradation efficiency of organic pollutants was substantially improved. ► Operation cost was greatly reduced compared to other microbial fuel cell designs. -- Abstract: Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H 2 O 2 generation and Fe 2+ release for the Fenton reaction. In tests using phenol, 75 ± 2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22 h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants

  1. Thermal annealing of an embrittled reactor pressure vessel

    International Nuclear Information System (INIS)

    Mager, T.R.; Dragunov, Y.G.; Leitz, C.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

  2. An Overview of the International Reactor Physics Experiment Evaluation Project

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Gulliford, Jim

    2014-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  3. Physics design of an ultra-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

    Ogawa, Y.; Inoue, N.; Wang, J.; Yamamoto, T.; Okano, K.

    1993-01-01

    A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

  4. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  5. MINIMARS: An attractive small tandem mirror fusion reactor

    International Nuclear Information System (INIS)

    Perkins, L.J.; Logan, B.G.; Doggett, J.N.; Devoto, R.S.

    1986-01-01

    Through the innovative design of a novel end plug scheme employing octopole MHD stabilization, the authors present the conceptual design of ''MINIMARS'', a small commercial fusion reactor based on the tandem mirror principle. The current baseline for MINIMARS has a net electric output of 600 MWe and they have configured the design for short construction times, factory-built modules, inherently safe blanket systems, and multiplexing in station sizes of ≅ 600-2400 MWe. They demonstrate that the compact octopole end cell provides a number of advantages over the more conventional quadrupole (yin-yang) end cell encountered in the MARS tandem mirror reactor study, and enables ignition to be achieved with much shorter central cell lengths. Accordingly, being economic in small sizes, MINIMARS provides an attractive alternative to the more conventional larger conceptual fusion reactors encountered to date, and would contribute significantly to the lowering of utility financial risk in a developing fusion economy

  6. An accelerator-driven reactor for meeting future energy demand

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi; Yang, Y.; Yu, A.

    1997-01-01

    Fissile fuel can be produced at a high rate using an accelerator-driven Pu-fueled subcritical fast reactor which avoids encountering a shortage of Pu during a high growth rate in the production of nuclear energy. Furthermore, the necessity of the early introduction of the fast reactor can be moderated. Subcritical operation provides flexible nuclear energy options along with high neutron economy for producing the fuel, for transmuting high-level waste such as minor actinides, and for efficiently converting excess and military Pu into proliferation-resistant fuel

  7. Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

    International Nuclear Information System (INIS)

    Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E.; Herring, J. Stephen

    2009-01-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered

  8. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  9. Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T.

    1997-01-01

    The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

  10. Effect of Fuel Structure Materials on Radiation Source Term in Reactor Core Meltdown

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Ha, Kwang Soon

    2014-01-01

    The fission product (Radiation Source) releases from the reactor core into the containment is obligatorily evaluated to guarantee the safety of Nuclear Power Plant (NPP) under the hypothetical accident involving a core meltdown. The initial core inventory is used as a starting point of all radiological consequences and effects on the subsequent results of accident assessment. Hence, a proper evaluation for the inventory can be regarded as one of the most important part over the entire procedure of accident analysis. The inventory of fission products is typically evaluated on the basis of the uranium material (e.g., UO2 and USi2) loaded in nuclear fuel assembly, except for the structure materials such as the end fittings, grids, and some kinds of springs. However, the structure materials are continually activated by the neutrons generated from the nuclear fission, and some nuclides of them (e.g., 14 C and 60 Co) can significantly influence on accident assessment. During the severe core accident, the structure components can be also melted with the melting points of temperature relatively lower than uranium material. A series of the calculation were performed by using ORIGEN-S module in SCALE 6.1 package code system. The total activity in each part of structure materials was specifically analyzed from these calculations. The fission product inventory is generally evaluated based on the uranium materials of fuel only, even though the structure components of the assembly are continually activated by the neutrons generated from the nuclear fission. In this study, the activation calculation of the fuel structure materials was performed for the initial source term assessment in the accident of reactor core meltdown. As a result, the lower end fitting and the upper plenum greatly contribute to the total activity except for the cladding material. The nuclides of 56 Mn, '5 1 Cr, 55 Fe, 58 Co, 54 Mn, and 60 Co are analyzed to mainly effect on the activity. This result

  11. Sourcing from an Enterprise Architecture Perspective

    DEFF Research Database (Denmark)

    Gøtze, John; Axél, Mette

    2013-01-01

    Enterprise architecture offers direct input in order to define the appropriate combination of people, components and services used to produce and deliver products and services. Sourcing, as the process of optimally organizing and procuring this inside and outside an organization, is therefore...... a central enterprise architecture concern. In this chapter we will discuss how enterprise architecture can benefit an enterprise’s sourcing strategy....

  12. Experimental investigation of thermal limits in parallel plate configuration for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; Kaminaga, M.; Yoder, G.L.

    1993-01-01

    The Advanced Neutron Source Reactor (ANSR) is currently being designed to become the world's highest-flux, steady-state, thermal neutron source for scientific experiments. Highly subcooled, heavy-water coolant flows vertically upward at a very high velocity of 25 m/s through parallel aluminum fuel-plates. The core has average and peak heat fluxes of 5.9 and 12 MW/m 2 , respectively. In this configuration, both flow excursion (FE) and true critical heat flux (CHF), represent potential thermal limitations. The availability of experimental data for both FE and true CHF at the conditions applicable to the ANSR is very limited. A Thermal Hydraulic Test Loop (THTL) facility was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of both thermal limits under the expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 14 MW/m 2 and a corresponding velocity range of 8 to 21 m/s. Both the exit pressure (1.7 MPa) and inlet temperature (45 degrees C) were maintained constant for these tests, while the loop was operated in a ''stiff''(constant flow) mode. Limited experiments were also conducted at 12 MW/m 2 using a ''soft'' mode (near constant pressure-drop) for actual FE burnout tests and using a ''stiff' mode for true CHF tests, to compare with the original FE experiments

  13. The High Flux Isotope Reactor (HFIR) cold source project at ORNL

    International Nuclear Information System (INIS)

    Selby, D.

    1998-01-01

    The scope of this project includes the development, design, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube hole in the pressure vessel. All aspects of the cold source design will be based on demonstrated technology adapted to the HFIR design and operating conditions

  14. An overview of thermalhydraulics R and D for SLOWPOKE heating reactors

    International Nuclear Information System (INIS)

    Dimmick, G.R.

    1988-09-01

    AECL is currently demonstrating the use of pool-type reactors of up to 10 MW output to produce hot water at about 90 degrees Celsius. The initial focus for the development is the provision of a source of hot water for institutional and municipal heating networks. Ongoing developments are designed to broaden the applications to electricity generation and industrial processes such as desalination and agricultural needs. The reactor concept is based on the Slowpoke-2 research reactor, eight of which are successfully operating in Canada and abroad. The primary-circuit flow is driven by natural convection, with the heated water, produced by the reactor core near the bottom of the pool, being ducted to low-pressure-drop heat exchangers in the upper part of the pool. As the pool volume is relatively large, the fluid transit time around the circuit is long, ensuring that the reactor response to all normal transients is extremely slow. To investigate thermalhydraulics aspects of the reactor design, including its behaviour underextreme conditions, an electrically heated, natural-convection loop was designed and constructed. The core of the loop consists of a rod bundle that is a precise reproduction of one quarter of the core of the 2-MW SLOWPOKE Demonstration Reactor presently being tested at the Whiteshell Nuclear Research Establishment. With this loop, measurements of the distribution of pressure, temperature, velocity and subcooled void have been made in the simulated core, via a variety of intrusive and non-intrusive techniques. In addition, both the single- and two-phase behaviour of the system have been studied. This paper gives examples of the various in-core measurements made and also makes comparisons between the measured system behaviour and that predicted by the various steady-state and transient computer codes

  15. EURAC: the JRC proposal for an European fusion reactor materials test and development facility

    International Nuclear Information System (INIS)

    Kley, W.; Bishop, G.R.

    1986-01-01

    For the last 7 years we examined the use of a Spallation Neutron Source (SNS) as an altenative European Option to FMIT. For an optimized spallation neutron source design we find now for the same beam power the following design parameters: - Linear Accelerator: 600 MeV, 6 m-A-proton beam on liquid lead target - irradiation parameters: 320 dpa/year in 20 cm 3 or 274 dpa/year in 31.5 cm 3 6 -1 sec -1 in order to simulate the Pulsed Mode of Tokamak Power Reactors. The deflected beam can be used for other experiments

  16. Anaerobic acidification of glucose in an upflow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zoetemeyer, R J; Matthijsen, A J.C.M.; Van den Heuvel, J C; Cohen, A; Boelhouwer, C

    1982-01-01

    Glucose (10 and 50 kg/cubic meters) was effectively biodegraded to mainly butyrate, lactate, and acetate in an anaerobic upflow reactor at minimum residence times of 26 and 82 minutes, respectivelyly. Sludge granulation occurred at residence times of 1 and 6 hours, respective, which increased biomass retention. Maximum glucose conversion was 450 and 630 kg/cubic meters-day.

  17. An analysis of CDTN performance in the reactors technology area

    International Nuclear Information System (INIS)

    Pinheiro, R.B.

    1985-01-01

    The author makes an analysis of CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) performance in the reactors technology area, showing difficulties and failures, but emphasizing the particular competence and capacity acquired in this area, as for example: the capacity in codes and methods are of neutronic calculations and nuclear projects, experimental thermohydraulic program, tests services in components and the others. (C.M.) [pt

  18. Processing test of an upgraded mechanical design for PERMCAT reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borgognoni, Fabio, E-mail: fabio.borgognoni@enea.i [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Demange, David; Doerr, Lothar [Forschungszentrum Karlsruhe GmbH, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany); Tosti, Silvano [Associazione ENEA-Euratom sulla Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati, Roma I-00044 (Italy); Welte, Stefan [Forschungszentrum Karlsruhe GmbH, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany)

    2010-12-15

    The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H{sub 2}O and D{sub 2}.

  19. Studying the effect of xenon poisoning on the power of the Syrian miniature neutron source reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.

    1999-07-01

    The uranium 235 is often used as a fuel to produce the energy in nuclear reactors. Uranium nuclei are fissioned with thermal neutrons and produce energy plus a number of neutrons. A fraction of such fission neutrons is involved in other fission with new nuclei to sustain the fission reactions. The remain fraction of the neutrons is lost from the reactor in two ways: escaped from the reactor, or absorbed with other nuclei that exist in the reactor before or produced from fission. Fission nuclei which absorb neutrons heavily are called p oison , such as Xe 135. Because Xe 135 absorbs neutrons heavily, it reduces the number of neutrons in the reactor. Hence, Xe 135 is studied explicitly in the MNSR reactor, and calculation of its negative reactivity is presented in this research during the operation, equilibrium, and after the shutting down of the reactor. (author)

  20. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  1. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

    Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

  2. Dosimetry issues for an ultra-high flux beam and multipurpose research reactor design

    International Nuclear Information System (INIS)

    West, C.D.

    1993-01-01

    The Advanced Neutron Source is a new user facility for all fields of neutron research, including neutron beam experiments, materials analysis, materials testing, and isotope production. The complement and layout of the experimental facilities have been determined sufficiently, at a conceptual design level, to make reliable cost and schedule estimates. The source of neutrons will be a heavy water reactor, constructed largely of aluminum, with an available thermal neutron flux 5--10 times higher than existing research reactors. Among the dosimetry issues to be faced are damage prediction and surveillance for component life attainment; measurement of fluence and spectra in regions where both change substantially over a distance of a few centimeters; and characterization and measurement of the radiation field in the research areas around the neutron beam experiments

  3. Economics of seawater desalination with innovative nuclear reactors and other energy sources: the EURODESAL project

    International Nuclear Information System (INIS)

    Nisan, S.; Volpi, L.

    2004-01-01

    This paper summarises our recent investigations undertaken as part of the EURODESAL project on nuclear desalination, which were carried out by a consortium of four EU and one Canadian, Industrials and two leading EU R and D organisations. Major results of the project, in particular of its economic evaluation work package as discussed in this paper, are: 1. A coherent demonstration of the technical feasibility of nuclear desalination through the development of technical principles for the optimum cogeneration of electricity and water and by exploring the unique capabilities of the innovative nuclear reactors and desalination technologies; verification that the integrated system design does not adversely affect nuclear reactor safety. 2. The development of codes and methods for an objective assessment of the competitiveness and sustainability of proposed solutions through comparison, in European conditions, with fossil and renewable energy based solutions. The results obtained so far seem to be quite encouraging as regards the economical viability of nuclear desalination options. Thus, for example, specific desalination costs ($/m 3 of desalted water) for nuclear systems such as the AP600 and the French PWR900 (reference base case), coupled to Multiple Effect Distillation (MED) or the Reverse Osmosis (RO) processes, are 30% to 60% lower than fossil energy based systems using pulverised coal and natural gas with combined cycle, at low discount rates and recommended fuel prices. Even in the most unfavourable scenarios for nuclear energy (discount rates = 10%, low fossil fuel prices) desalination costs with the nuclear options with the nuclear reactors are 7% to 15% lower, depending upon the desalination capacities. Furthermore, with the high performance coupling schemes developed by the EURODESAL partners, the specific desalination costs of nuclear systems are reduced by another 2% to 14%, even without system and design optimisation. (author)

  4. Ultrafast electron diffraction using an ultracold source

    Directory of Open Access Journals (Sweden)

    M. W. van Mourik

    2014-05-01

    Full Text Available The study of structural dynamics of complex macromolecular crystals using electrons requires bunches of sufficient coherence and charge. We present diffraction patterns from graphite, obtained with bunches from an ultracold electron source, based on femtosecond near-threshold photoionization of a laser-cooled atomic gas. By varying the photoionization wavelength, we change the effective source temperature from 300 K to 10 K, resulting in a concomitant change in the width of the diffraction peaks, which is consistent with independently measured source parameters. This constitutes a direct measurement of the beam coherence of this ultracold source and confirms its suitability for protein crystal diffraction.

  5. The measurement of subcritical reactivity in nuclear reactors by use of a high frequency sine-wave modulated neutron source

    International Nuclear Information System (INIS)

    Guppy, C.B.

    1964-11-01

    In this report the frequency response characteristics for phase and gain of the fundamental reactor mode of the zero power kinetics are given for various subcritical reactivities in a fast reactor and in a thermal reactor. Results, of a study on harmonic effects based on a small zero energy thermal reactor are presented which demonstrate the importance of spatial harmonic effects. A harmonic theory for thermal reactors is developed. A new method of measuring, subcritical reactivity at moderately high frequencies is suggested which circumvents the harmonic problem. It is shown that at high frequencies there is more sensitivity than at low frequencies and that this could lead to an increased range over which subcritical reactivity can be measured. (author)

  6. A station blackout simulation for the Advanced Neutron Source Reactor using the integrated primary and secondary system model

    International Nuclear Information System (INIS)

    Schneider, E.A.

    1994-01-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at Oak Ridge National Laboratory. This paper deals with thermal-hydraulic analysis of ANSR's cooling systems during nominal and transient conditions, with the major effort focusing upon the construction and testing of computer models of the reactor's primary, secondary and reflector vessel cooling systems. The code RELAP5 was used to simulate transients, such as loss of coolant accidents and loss of off-site power, as well as to model the behavior of the reactor in steady state. Three stages are involved in constructing and using a RELAP5 model: (1) construction and encoding of the desired model, (2) testing and adjustment of the model until a satisfactory steady state is achieved, and (3) running actual transients using the steady-state results obtained earlier as initial conditions. By use of the ANSR design specifications, a model of the reactor's primary and secondary cooling systems has been constructed to run a transient simulating a loss of off-site power. This incident assumes a pump coastdown in both the primary and secondary loops. The results determine whether the reactor can survive the transition from forced convection to natural circulation

  7. Neutron energy spectrum flux profile of Ghana's miniature neutron source reactor core

    International Nuclear Information System (INIS)

    Sogbadji, R.B.M.; Abrefah, R.G.; Ampomah-Amoako, E.; Agbemava, S.E.; Nyarko, B.J.B.

    2011-01-01

    Highlights: → The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was studied. → Using 20,484 energy grids, the thermal, slowing down and fast neutron energy regions were studied. - Abstract: The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) x 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) x 10 11 n/cm 2 s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) x 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939-3.7880) x 10 -08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) x 10 09 n/cm 2 s at the lower energy end of the slowing down region between 8.2491 x 10 -01 MeV and 8.2680 x 10 -01 MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) x 10 08 n/cm 2 s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.

  8. Power balance in an Ohmically heated fusion reactor

    International Nuclear Information System (INIS)

    Christiansen, J.P.; Roberts, K.V.

    1982-01-01

    A simplified power-balance equation (zero-dimensional model) is used to study the performance of an Ohmically heated fusion reactor with emphasis on a pulsed reversed-field pinch concept (RFP). The energy confinement time tausub(E) is treated as an adjustable function, and empirical tokamak scaling laws are employed in the numerical estimates, which are supplemented by 1-D ATHENE code calculations. The known heating rates and energy losses are represented by the net energy replacement time tausub(W), which is exhibited as a surface in density (n) and temperature (T) space with a saddle point (nsub(*), Tsub(*)), the optimum ignition point. It is concluded that i) ignition by Ohmic heating is more practicable for the RFP reactor than for a tokamak reactor with the same tausub(E), (ii) if at fixed current the minor radius can be reduced or at fixed minor radius the current can be increased, then it is found that Ohmic ignition becomes more likely when present tokamak scaling laws are used. More definitive estimates require, however, a knowledge of tausub(E), which can only be obtained by establishing a reliable set of experimental RFP scaling laws and, in particular, by extending RFP experiments closer to the reactor regime. (author)

  9. Towards an efficient multiphysics model for nuclear reactor dynamics

    Directory of Open Access Journals (Sweden)

    Obaidurrahman K.

    2015-01-01

    Full Text Available Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient.

  10. Development of an automated core model for nuclear reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input

  11. Analysis of an homogeneous solution reactor for 99 Mo production

    International Nuclear Information System (INIS)

    Weir, A.; Lopasso, E.; Gho, C.

    2007-01-01

    The 99m Tc is the more used radioisotope in nuclear medicine, used in 80% of procedures of nuclear medicine in the world. This is due to their characteristics practically ideal for the diagnostic. The 99m Tc is obtained by decay of the 99 Mo, which can produce it by irradiating enriched targets in 98 Mo, or as fission product, irradiating uranium targets or by means of homogeneous solution reactors. The pattern of the used reactor in the neutron analysis possesses a liquid fuel composed of uranyl nitrate dissolved in water with the attach of nitric acid. This solution is contained in a cylindrical recipient of stainless steel reflected with light water. The reactor is refrigerated by means of an helicoidal heat exchanger immersed in the fuel solution. The heat of the fuel is removed by natural convection while the circulation of the water inside the exchanger is forced. The control system of the reactor consists on 6 independent cadmium bars, with followers of water. An auxiliary control system can be the level of the fuel solution inside container tank, but it was not included in the pattern in study. One studies the variations of the reactivity of the system due to different phenomena. An important factor during the normal operation of the reactor is the variation of temperature taking to a volumetric expansion of the fuel and ghastly effects in the same one. Another causing phenomenon of changes in the reactivity is the variation of the concentration of uranium in the combustible solution. An important phenomenon in this type of reactors is the hole fraction in the nucleus I liquidate due to the radiolysis and the possible boil of the water of the combustible solution. Some of the possible cases of abnormal operation were studied as the lost one of coolant in the secondary circuit of the heat exchanger, the introduction and evaporation of water in the nucleus. The reactivity variations were studied using the codes of I calculate MCNP, WIMS and TORT. All the

  12. Modifications to ORIGEN2 for generating N Reactor source terms. Volume 3: ORIGEN2 N-Reactor output files

    International Nuclear Information System (INIS)

    1997-01-01

    This text is intended to be a brief outline of the ORIGEN2 computer code which is a revised and updated version of the ORIGEN documented in report ORNL-4628 (May 1973). Included here are: a brief description of the functions of ORIGEN2; a listing of the major data sources; a listing of the published documentation concerning ORIGEN2; and an outline of the ORIGEN2 output organization. ORIGEN2 is available from the ORNL Radiation Shielding Information Center (RSIC). Past experience has indicated that many users encounter considerable difficulty in finding the desired information in a ORIGEN2 output which is sometimes rather massive. This section is intended as a brief outline of the organization of ORIGEN2 output

  13. A model for reliability avaliation of the electrical supply source of the 1A3 and 1A4 control rods assemblies of Angra I reactor

    International Nuclear Information System (INIS)

    Yang, T.

    1978-01-01

    The reliability of the electrical power supply to the 4.16KV buses for the safety system operation of a nuclear power plant was studied. Particularly, Angra Unit I system was focused. Initially, reliability of each electrical supply source was estimated. Using a probabilistic approach based on the Markov processes, the system reliability was evaluated in terms of frequency and duration of loss of power supply and of the system failure probability evolution when one or more sources remained unavailable. Based on these results, certain reactor operating rules were proposed concerning later shutdown of the plant without compromising the nuclear reactor safety. A sensitivity analysis was also performed to show the different reliability parameter influences on final results. This analysis showed that the diesel system performs an important role in the power supply for a nuclear power plant [pt

  14. Neutron activation analysis in an industrial laboratory using an off-site nuclear reactor

    International Nuclear Information System (INIS)

    Osborn, T.W.; Broering, W.B.

    1977-01-01

    A multifunctional research laboratory, such as Procter and Gamble's Miami Valley Laboratories, requires elemental analyses on many materials. A general survey technique is important even if the information it provides is incomplete or is less precise than single element analyses. Procter and Gamble has developed neutron activation analysis (NAA) capabilities using a nuclear reactor several hundred miles away. The concentration of 40 to 50 elements can be determined in a variety of matrices. We have found NAA to be a powerful supplement to some of the more classical analytical techniques even without having an on-site neutron source. We have also found an automated data acquisition system to be essential for the successful application of NAA in an industrial laboratory

  15. Modernization of the High Flux Isotope Reactor (HFIR) to Provide a Cold Neutron Source and Experimentation Facility

    International Nuclear Information System (INIS)

    Rothrock, Benjamin G.; Farrar, Mike B.

    2009-01-01

    In June 1961, construction was started on the High Flux Isotope Reactor (HFIR) facility inside the Oak Ridge National Laboratory (ORNL), at the recommendation of the U.S. Atomic Energy Commission (AEC) Division of Research. Construction was completed in early 1965 with criticality achieved on August 25, 19651. From the first full power operating cycle beginning in September 1966, the HFIR has achieved an outstanding record of service to the scientific community. In early 1995, the ORNL deputy director formed a group to examine the need for upgrades to the HFIR following the cancellation of the Advanced Neutron Source Project by DOE. This group indicated that there was an immediate need for the installation of a cold neutron source facility in the HFIR to produce cold neutrons for neutron scattering research uses. Cold neutrons have long wavelengths in the range of 4-12 angstroms. Cold neutrons are ideal for research applications with long length-scale molecular structures such as polymers, nanophase materials, and biological samples. These materials require large scale examination (and therefore require a longer wavelength neutron). These materials represent particular areas of science are at the forefront of current research initiatives that have a potentially significant impact on the materials we use in our everyday lives and our knowledge of biology and medicine. This paper discusses the installation of a cold neutron source at HFIR with respect to the project as a modernization of the facility. The paper focuses on why the project was required, the scope of the cold source project with specific emphasis on the design, and project management information.

  16. RELAP5 analyses of two hypothetical flow reversal events for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1995-01-01

    The reactor design features 4 independent cooling loops (3 active, 1 standby), each containing a main circulation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and also that the check valve in that loop remains stuck open. This accident is considered extremely unlikely. Flow reverses in this loop, reducing core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-dia instantaneous pipe break near the core inlet (worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against 4 thermal limits: T wall = T sat , incipient boiling, onset of significant void, and critical heat flux. For the first transient, results show that these limits are not exceeded (at 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (present design value). For the second transient, the closest approach of the fuel surface temperature to local saturation temperature during core flow reversal is about 39 C, so the fuel remains cool during this transient. Although this work is for the ANSR geometry and operating conditions, the general conclusion may be applicable to other highly subcooled reactor systems

  17. Reactor materials research as an effective instrument of nuclear reactor perfection

    International Nuclear Information System (INIS)

    Baryshnikov, M.

    2006-01-01

    The work is devoted to reactor materiology, as to the practical tool of nuclear reactor development. The work is illustrated with concrete examples from activity experience of the appropriate division of the Russian Research Centre Kurchatov Institute - Institute of Reactor Materials Research and Radiation Nanotechnologies. Besides the description of some modern potentials of the mentioned institute is given. (author)

  18. Treatment of pig excreta using an SCFBR anaerobic reactor

    Directory of Open Access Journals (Sweden)

    Kevin G. Molina T.

    1999-01-01

    Full Text Available A new anaerobic reactor called the Sludge Central Fixed Bed Reactor (SCFBR was built and evaluated for the treatment of liquid residue from the pig farms. The SCFBR has three main parts. The lower area is for sludge, the middle part consists of a concentrically packed zone and an upper area for the separation of solids, liquids and gases. The 28.51 SCFBR reactor was evaluated over a period of 210 days, using three organic loads of 0.548,0.421 and 1.239 g COD/1 day. Initially, the reactor was fed non-continuously using 10 and 10.7 days Hydraulic Retention Times (HRT. The HRT was later decreased to 3.87 days using continuous feeding. For the three 0.548, 0.421 and 1.239 g COD/1 day organic loads, Chemical Oxygen Demand (COD removal was 68%, 81% and 73% respectively and Volatile Solids (VS removal was 53.5%, 55.8% and 50.1% respectively. The SCFBR performed well, as shown by the removal efficiency and stability obtained. A microphotograph of sludge from the lower zone is presented, showing high Methanosaeta (Methanothrix presence.

  19. Real-time numerical simulation with high efficiency for an experimental reactor system

    International Nuclear Information System (INIS)

    Ding Shuling; Li Fu; Li Sifeng; Chu Xinyuan

    2006-01-01

    The paper presents a systematic and efficient method for numerical real-time simulation of an experimental reactor. The reactor models were built based on the physical characteristics of the experimental reactor, and several real-time simulation approaches were discussed and compared in the paper. How to implement the real-time reactor simulation system in Windows platform for the sake of hardware-in-loop experiment for the reactor power control system was discussed. (authors)

  20. System and method for air temperature control in an oxygen transport membrane based reactor

    Science.gov (United States)

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  1. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1994-01-01

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  2. An Overview of the Safety Case for Small Modular Reactors

    International Nuclear Information System (INIS)

    Ingersoll, Daniel T.

    2011-01-01

    Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.

  3. Design of a homogeneous subcritical nuclear reactor based on thorium with a source of californium 252; Diseno de un reactor nuclear subcritico homogeneo a base de Torio con una fuente de Californio 252

    Energy Technology Data Exchange (ETDEWEB)

    Delgado H, C. E.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: ce_delgado89@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of)

    2015-10-15

    Full text: One of the energy alternatives to fossil fuels which do not produce greenhouse gases is the nuclear energy. One of the drawbacks of this alternative is the generation of radioactive wastes of long half-life and its relation to the generation of nuclear materials to produce weapons of mass destruction. An option to these drawbacks of nuclear energy is to use Thorium as part of the nuclear fuel which it becomes in U{sup 233} when capturing neutrons, that is a fissile material. In this paper Monte Carlo methods were used to design a homogeneous subcritical reactor based on thorium. As neutron reflector graphite was used. The reactor core is homogeneous and is formed of 70% light water as moderator, 12% of enriched uranium UO{sub 2}(NO{sub 3}){sub 4} and 18% of thorium Th(NO{sub 3}){sub 4} as fuel. To start the nuclear fission chain reaction an isotopic source of californium 252 was used with an intensity of 4.6 x 10{sup 7} s{sup -1}. In the design the value of the effective multiplication factor, whose value turned out k{sub eff} <1 was calculated. Also, the neutron spectra at different distances from the source and the total fluence were calculated, as well as the values of the ambient dose equivalent in the periphery of the reactor. (Author)

  4. Biomass as an alternative energy source

    Energy Technology Data Exchange (ETDEWEB)

    De Bruyn, M.; Naveau, H.; Declerck, C.; Vanacker, L.; Mahy, D.; Schepens, G.

    The object of this paper is to evaluate the possible production and utilization of biomass as an energy source in Belgium. Four conversion methods are considered - methanation, fermentation, incineration and gasification - from a technological and economic viewpoint.

  5. Natural circulation analysis for the advanced neutron source reactor refueling process 11

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, R.F.; Dasardhi, S.; Elkassabgi, Y. [Texas A& M Univ., Kingsville, TX (United States); Yoder, G.L. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    During the refueling process of the Advanced Neutron Source Reactor (ANSR), the spent fuel elements must be moved from the primary coolant loop (containing D{sub 2}O), through a heavy water pool, and finally into a light water spent fuel storage area. The present refueling scheme utilizes remote refueling equipment to move the spent fuel elements through a D{sub 2}O filled stack and tunnel into a temporary storage canal. A transfer lock is used to move the spent fuel elements from the D{sub 2}O-filled interim storage canal to a light water pool. Each spent fuel element must be cooled during this process, using either natural circulation or forced convection. This paper presents a summary of the numerical techniques used to analyze natural circulation cooling of the ANSR fuel elements as well as selected results of the calculations. Details of the analysis indicate that coolant velocities below 10 cm/s exist in the coolant channels under single phase natural circulation conditions. Also, boiling does not occur within the channels if power levels are below a few hundred kW when the core transitions to natural circulation conditions.

  6. Seismic explosion sources on an ice cap

    DEFF Research Database (Denmark)

    Shulgin, Alexey; Thybo, Hans

    2015-01-01

    crustal model can be modelled. A crucial challenge for applying the technique is to control the sources. Here, we present data that describe the efficiency of explosive sources in the ice cover. Analysis of the data shows, that the ice cap traps a significant amount of energy, which is observed......Controlled source seismic investigation of crustal structure below ice covers is an emerging technique. We have recently conducted an explosive refraction/wide-angle reflection seismic experiment on the ice cap in east-central Greenland. The data-quality is high for all shot points and a full...

  7. Tokamak Fusion Test Reactor - CICADA, an overview

    International Nuclear Information System (INIS)

    Sauthoff, N.; Bosco, J.; Daniels, R.

    1983-01-01

    The Central Instrumentation, Control and Data Acquisition (CICADA) System provides the interface between the human operators and the equipment to be controlled and monitored in support of TFTR operation. The functions can be partitioned into those that are instantaneous and fundamental components of a processed control system (such as displaying the current monitored value of a point or changing the state of a piece of equipment) and those that are more complex, being composite actions executed in sequences either related to the TFTR shot cycle or in the execution of timed procedures for the operation of subsystems. In this paper, the authors present the configuration of the equipment ''from the outside in'' - meaning from the man and equipment external interfaces inward toward the components internal to the CICADA system; they next present an overview of the fundamental operations, including the concept of the CICADA devices, the assignment of values to the elements of the definition of the device, the monitoring procedure for all of the devices on the system, the control process including the assignment of control authorization, the handling of special devices such as diagnostic data acquisition and timing modules, and graphics. Next they proceed to discuss composite actions and timed sequences which support the TFTR cycle; included are the system for activating, suspending, and resuming tasks in response to the occurrence of hardware and software events, and the typical sequence of tasks involved in operation of a subsystem for data acquisition during the shot cycle and the archival of raw and results data; special cases will be the real time control systems for the plasma position and current and the gas injection system, both of which provide closed-loop feedback control of plasma position, current and lineintegrated density throughout the shot. Finally, future directions and initiatives discussed

  8. An ideal sealed source life-cycle

    International Nuclear Information System (INIS)

    Tompkins, Joseph Andrew

    2009-01-01

    we have today. This regulation created a new regulatory framework seen as promising at the time. However, now they recognize that, despite the good intentions, the NIJWP/85 has not solved any source disposition problems. The answer to these sealed source disposition problems is to adopt a philosophy to correct these regulatory issues, determine an interim solution, execute that solution until there is a minimal backlog of sources to deal with, and then let the mechanisms they have created solve this problem into the foreseeable future. The primary philosophical tenet of the ideal sealed source life cycle follows. You do not allow the creation (or importation) of any source whose use cannot be justified, which cannot be affordably shipped, or that does not have a well-delinated and affordable disposition pathway. The path forward dictates that we fix the problem by embracing the Ideal Source Life cycle. In figure 1, we can see some of the elements of the ideal source life cycle. The life cycle is broken down into four portions, manufacture, use, consolidation, and disposition. These four arbitrary elements allow them to focus on the ideal life cycle phases that every source should go through between manufacture and final disposition. As we examine the various phases of the sealed source life cycle, they pick specific examples and explore the adoption of the ideal life cycle model.

  9. Defining an Open Source Strategy for NASA

    Science.gov (United States)

    Mattmann, C. A.; Crichton, D. J.; Lindsay, F.; Berrick, S. W.; Marshall, J. J.; Downs, R. R.

    2011-12-01

    Over the course of the past year, we have worked to help frame a strategy for NASA and open source software. This includes defining information processes to understand open source licensing, attribution, commerciality, redistribution, communities, architectures, and interactions within the agency. Specifically we held a training session at the NASA Earth Science Data Systems Working Group meeting in Open Source software as it relates to the NASA Earth Science data systems enterprise, including EOSDIS, the Distributed Active Archive Centers (DAACs), ACCESS proposals, and the MEASURES communities, and efforts to understand how open source software can be both consumed and produced within that ecosystem. In addition, we presented at the 1st NASA Open Source Summit (OSS) and helped to define an agency-level strategy, a set of recommendations and paths forward for how to identify healthy open source communities, how to deal with issues such as contributions originating from other agencies, and how to search out talent with the right skills to develop software for NASA in the modern age. This talk will review our current recommendations for open source at NASA, and will cover the set of thirteen recommendations output from the NASA Open Source Summit and discuss some of their implications for the agency.

  10. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  11. Feasibility studies on the production of essential radioisotopes (24Na and 32P) using the Ghana Miniature Neutron Source Reactor (GHARR-1)

    International Nuclear Information System (INIS)

    Dotse, S.C.

    2012-01-01

    Feasibility studies on the production of 32 P and 24 Na using a Miniature Neutron Source Reactor named Ghana Research Reactor-1 (GHARR-1) has been conducted. A theoretical model of the cyclic activation technique was developed for the simulation of specific activities under varying parameters. Specific activity values estimated for four cycles of irradiation with activation parameters falling within the specifications of the reactor were experimentally validated. Experimented results were compared to those theoretically estimated for both 24 Na and 32 P. Experimented specific activity values for both radioisotopes generally fell below their theoretical values but recorded activity build-ups from one cycle to the other. The 24 Na nuclide showed a regular pattern for the increase in specific activity from one cycle to the next with an average percentage difference of experimented to theoretical value of 19%. The 32 P nuclide showed an irregular pattern for the increase in specific activity from one cycle to the next with an average percentage difference of experimental to theoretical value of 11%. The specific activities experimentally attained, with reference to activity levels used for various applications in agriculture and industry suggests the cyclic activation technique can be used for the production of radioisotopes of appreciable activities using low power research reactors, which are characterised with limited excess core reactivity and cannot sustain long periods of irradiation. (au)

  12. Advanced Neutron Source (ANS) Project Progress report, FY 1991

    International Nuclear Information System (INIS)

    Campbell, J.H.; Selby, D.L.; Harrington, R.M.; Thompson, P.B.

    1992-01-01

    This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown Concepts; Critical and Subcritical Experiments; Material Data, Structural Tests, and Analysis; Cold-Source Development; Beam Tube, Guide, and Instrument Development; Hot-Source Development; Neutron Transport and Shielding; I ampersand C Research and Development; Design; and Safety

  13. An holistic approach to the problem of reactor ageing

    International Nuclear Information System (INIS)

    Phythian, W.; McElroy, R.; Druce, S.; Kovan, D.

    1992-01-01

    Understanding the process of ageing in reactors is essential to extending their lives beyond original design. To present a sound case -particularly regarding the level of embrittlement in reactor vessels due to radiation damage - an integrated approach using advanced assessment tools is needed. The techniques developed for the purpose involve, on the microscopic level, advanced neutron dosimetry and high resolution measurement techniques (eg advanced electron beam techniques and small angle neutron scattering) with which an analysis can be done of the radiation damage and the microstructural state of the steel test procedures (tensile, fracture toughness and Charpy impact) on standard and sub-sized specimens, the extent of radiation degradation can be characterised. finally, it is possible to predict how the degradation will evolve using physically-based models of embrittlement. (Author)

  14. Chemical modification and blending of polymers in an extruder reactor

    International Nuclear Information System (INIS)

    Prut, Eduard V; Zelenetskii, Alexandr N

    2001-01-01

    Chemical modification and blending of polymers in an extruder reactor are discussed. Relationships between the parameters affecting the reaction kinetics, viz., mixing time, duration of a chemical reaction and the residence time of the system in the extruder reactor, and the structure of the materials produced are analysed. The mechanisms of (i) grafting of low-molecular-mass compounds onto polymers; (ii) reactions between terminal groups of different polymers and (iii) transesterification and interchange reactions are considered. The factors affecting the mechanism of dynamic vulcanisation and the properties of thermoplastic elastomers are identified. Solid-phase reactions of polysaccharides in an extruder are discussed. The priority aspects of studies on the chemical modification and blending of polymers are noted. The bibliography includes 90 references.

  15. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  16. A review of source term and dose estimation for the TMI-2 reactor accident

    International Nuclear Information System (INIS)

    Gudiksen, P.H.; Dickerson, M.H.

    1990-09-01

    The TMI-2 nuclear reactor accident, which occurred on March 28, 1979 in Harrisburg, Pennsylvania, produced environmental releases of noble gases and small quantities of radioiodine. The releases occurred over a roughly two week period with almost 90% of the noble gases being released during the first three days after the initiation of the accident. Meteorological conditions during the prolonged release period varied from strong synoptic driven flows that rapidly transported the radioactive gases out of the Harrisburg area to calm situations that allowed the radioactivity to accumulate within the low lying river area and to subsequently slowly disperse within the immediate vicinity of the reactor. The results reported by various analysts, revealed that approximately 2.4--10 million curies of noble gases (mainly Xe-133), and about 14 curies of I-131 were released. During the first two days, when most of the noble gas release occurred, the plume was transported in a northerly direction causing the most exposed area to lie within a northwesterly to northeasterly direction from TMI. Changing surface winds caused the plume to be subsequently transported in a southerly direction, followed by an easterly direction. The calculated maximum whole body dose due to plume passage exceeded 100 mrem over an area extending several kilometers north of the plant, although the highest measured dose was 75 mrem. The collective dose equivalent (within a radius of 80 km) due to the noble gas exposure ranged over several orders of magnitude with a central estimate of 3300 person-rem. The small I-131 release produced barely detectable levels of activity in air and milk samples. This may have produced thyroid doses of a few milirem to a small segment of the population. 7 refs., 4 figs., 2 tabs

  17. The Effect of Initial Inoculum Source on the Microbial Community Structure and Dynamics in Laboratory-Scale Sequencing Batch Reactors

    KAUST Repository

    Hernandez, Susana

    2011-07-01

    Understanding the factors that shapes the microbial community assembly in activated sludge wastewater treatment processes provide a conceptual foundation for improving process performance. The aim of this study was to compare two major theories (deterministic theory and neutral theory) regarding the assembly of microorganisms in activated sludge: Six lab-scale activated sludge sequencing batch reactors were inoculated with activated sludge collected from three different sources (domestic, industrial, and sugar industry WWTP). Additionally, two reactors were seeded with equal proportion of sludge from the three WWTPs. Duplicate reactors were used for each sludge source (i.e. domestic, industrial, sugar and mix). Reactors were operated in parallel for 11 weeks under identical conditions. Bacterial diversity and community structure in the eight SBRs were assessed by 16S rRNA gene pyrosequencing. The 16S rRNA gene sequences were analyzed using taxonomic and clustering analysis and by measuring diversity indices (Shannon-weaver and Chao1 indices). Cluster analysis revealed that the microbial community structure was dynamic and that replicate reactors evolved differently. Also the microbial community structure in the SBRs seeded with a different sludge did not converge after 11 weeks of operation under identical conditions. These results suggest that history and distribution of taxa in the source inoculum were stronger regulating factors in shaping bacterial community structure than environmental factors. This supports the neutral theory which states that the assembly of the local microbial community from the metacommunity is random and is regulated by the size and diversity of the metacommunity. Furthermore, sludge performance, measured by COD and ammonia removal, confirmed that broad-scale functions (e.g. COD removal) are not influenced by dynamics in the microbial composition, while specific functions (e.g. nitrification) are more susceptible to these changes.

  18. Study of cold neutron sources: Implementation and validation of a complete computation scheme for research reactor using Monte Carlo codes TRIPOLI-4.4 and McStas

    International Nuclear Information System (INIS)

    Campioni, Guillaume; Mounier, Claude

    2006-01-01

    The main goal of the thesis about studies of cold neutrons sources (CNS) in research reactors was to create a complete set of tools to design efficiently CNS. The work raises the problem to run accurate simulations of experimental devices inside reactor reflector valid for parametric studies. On one hand, deterministic codes have reasonable computation times but introduce problems for geometrical description. On the other hand, Monte Carlo codes give the possibility to compute on precise geometry, but need computation times so important that parametric studies are impossible. To decrease this computation time, several developments were made in the Monte Carlo code TRIPOLI-4.4. An uncoupling technique is used to isolate a study zone in the complete reactor geometry. By recording boundary conditions (incoming flux), further simulations can be launched for parametric studies with a computation time reduced by a factor 60 (case of the cold neutron source of the Orphee reactor). The short response time allows to lead parametric studies using Monte Carlo code. Moreover, using biasing methods, the flux can be recorded on the surface of neutrons guides entries (low solid angle) with a further gain of running time. Finally, the implementation of a coupling module between TRIPOLI- 4.4 and the Monte Carlo code McStas for research in condensed matter field gives the possibility to obtain fluxes after transmission through neutrons guides, thus to have the neutron flux received by samples studied by scientists of condensed matter. This set of developments, involving TRIPOLI-4.4 and McStas, represent a complete computation scheme for research reactors: from nuclear core, where neutrons are created, to the exit of neutrons guides, on samples of matter. This complete calculation scheme is tested against ILL4 measurements of flux in cold neutron guides. (authors)

  19. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  20. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  1. Degradation of whey in an anaerobic fixed bed (AnFB) reactor

    OpenAIRE

    Handajani, Marisa

    2004-01-01

    An Anaerobic Fixed Bed (AnFB) reactor was run as an upflow anaerobic reactor with an arrangement of supporting material for growth of a biofilm. The supporting material was made from Liapor-clay-polyethylene sinter lamellas (Herding Co., Amberg). The AnFB reactor was used for treating high concentrations of whey-containing wastewater. Optimal operating conditions for whey treatment at a concentration of COD in the influent of around 50 g whey·l-1 were found for a hydraulic retention ...

  2. Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L., E-mail: mabs500@gmail.com, E-mail: narain@ufpe.br, E-mail: jairbezerra@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociências. Departamento de Energia Nuclear

    2017-07-01

    Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

  3. Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

    International Nuclear Information System (INIS)

    Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L.

    2017-01-01

    Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

  4. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  5. Design requirement for electrical system of an advanced research reactor

    International Nuclear Information System (INIS)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S.

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system

  6. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  7. Plasma Ion Source with an Internal Evaporator

    International Nuclear Information System (INIS)

    Turek, M.; Drozdziel, A.; Pyszniak, K.; Prucnal, S.; Maczka, D.

    2011-01-01

    A new construction of a hollow cathode ion source equipped with an internal evaporator heated by a spiral cathode filament and arc discharge is presented. The source is especially suitable for production of ions from solids. The proximity of arc discharge region and extraction opening enables production of intense ion beams even for very low discharge current (I a = 1.2 A). The currents of 50 μA (Al + ) and 70 μA (Bi + ) were obtained using the extraction voltage of 25 kV. The source is able to work for several tens of hours without maintenance breaks, giving possibility of high dose implantations. The paper presents the detailed description of the ion source as well as its experimental characteristics like dependences of extracted currents and anode voltage on anode and cathode currents. (author)

  8. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    Energy Technology Data Exchange (ETDEWEB)

    Prabhudharwadkar, Deoras M. [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Iyer, Kannan N., E-mail: kiyer@iitb.ac.i [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Mohan, Nalini; Bajaj, Satinder S. [Nuclear Power Corporation of India Ltd., Mumbai (India); Markandeya, Suhas G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India)

    2011-03-15

    Research highlights: This work addresses hydrogen dispersion in commercial nuclear reactor containment. The numerical tool used for simulation is first benchmarked with experimental data. Parametric results are then carried out for different release configurations. Results lead to the conclusion that the dispersal is buoyancy dominated. Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is quite poor and most

  9. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    International Nuclear Information System (INIS)

    Prabhudharwadkar, Deoras M.; Iyer, Kannan N.; Mohan, Nalini; Bajaj, Satinder S.; Markandeya, Suhas G.

    2011-01-01

    Research highlights: → This work addresses hydrogen dispersion in commercial nuclear reactor containment. → The numerical tool used for simulation is first benchmarked with experimental data. → Parametric results are then carried out for different release configurations. → Results lead to the conclusion that the dispersal is buoyancy dominated. → Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is

  10. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1980-01-01

    The law intends under the principles of the atomic energy act to regulate the refining, processing and reprocessing businesses of nuclear raw and fuel metarials and the installation and operation of reactors for the peaceful and systematic utilization of such materials and reactors and for securing public safety by preventing disasters, as well as to control internationally regulated things for effecting the international agreements on the research, development and utilization of atomic energy. Basic terms are defined, such as atomic energy; nuclear fuel material; nuclear raw material; nuclear reactor; refining; processing; reprocessing; internationally regulated thing. Any person who is going to engage in refining businesses other than the Power Reactor and Nuclear Fuel Development Corporation shall get the special designation by the Prime Minister and the Minister of International Trade Industry. Any person who is going to engage in processing businesses shall get the particular admission of the Prime Minister. Any person who is going to establish reactors shall get the particular admission of the Prime Minister, The Minister of International Trade and Industry or the Minister of Transportation according to the kinds of specified reactors, respectively. Any person who is going to engage in reprocessing businesses other than the Power Reactor and Nuclear Fuel Development Corporation and the Japan Atomic Energy Research Institute shall get the special designation by the Prime Minister. The employment of nuclear fuel materials and internationally regulated things is defined in detail. (Okada, K.)

  11. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.

    1983-01-01

    In order to assure the continued utilization of fission energy, development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed as the best type for future Brazilian nuclear systems. The inherent safety characteristics are superior to current FBRs and an efficient utilization of the abundant thorium is possible. A first step and a basic tool for the development of FBR technologies is the construction and operation of an experimental fast reactor (EFR). A series of core designs for a 90 MW EFR is studied and possible options and the magnitudes of principal parameters are identified. Flexible modifications of the core and sufficiently high fast fluxes for fuel and materials irradiations appear possible. (Author) [pt

  12. Continuous removal and recovery of tellurium in an upflow anaerobic granular sludge bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mal, Joyabrata, E-mail: joyabrata2006@gmail.com [UNESCO-IHE, Westvest 7, 2611 AX Delft (Netherlands); Nancharaiah, Yarlagadda V. [Biofouling and Biofilm Processes Section, Water and Steam Chemistry Division, Bhabha Atomic Research Centre, Kalpakkam, 603102, Tamil Nadu (India); Homi Bhabha National Institute, Anushakti Nagar Complex, Mumbai 400094 (India); Maheshwari, Neeraj [CNRS UMR 7338, BMBI University de Technologie Compiegne, 60200 Compiegne (France); Hullebusch, Eric D. van [UNESCO-IHE, Westvest 7, 2611 AX Delft (Netherlands); Université Paris-Est, Laboratoire Géomatériaux et Environnement (LGE), EA 4508, UPEM, 77454, Marne-la-Vallée (France); Lens, Piet N.L. [UNESCO-IHE, Westvest 7, 2611 AX Delft (Netherlands); Department of Chemistry and Bioengineering, Tampere University of Technology, P.O-Box 541, Tampere (Finland)

    2017-04-05

    Highlights: • Tellurite bioreduction coupled to recovery of biogenic Te(0) nanocrystals. • First report on continuous tellurite removal in a UASB reactor. • Biogenic Te(0) was mainly associated with loosely-bound EPS of granular sludge. • Repeated exposure to tellurite caused compositional changes in the EPS matrix. - Abstract: Continuous removal of tellurite (TeO{sub 3}{sup 2−}) from synthetic wastewater and subsequent recovery in the form of elemental tellurium was studied in an upflow anaerobic granular sludge bed (UASB) reactor operated at 30 °C. The UASB reactor was inoculated with anaerobic granular sludge and fed with lactate as carbon source and electron donor at an organic loading rate of 0.6 g COD L{sup −1} d{sup −1}. After establishing efficient and stable COD removal, the reactor was fed with 10 mg TeO{sub 3}{sup 2−} L{sup −1} for 42 d before increasing the influent concentration to 20 mg TeO{sub 3}{sup 2−} L{sup −1}. Tellurite removal (98 and 92%, respectively, from 10 and 20 mg Te L{sup −1}) was primarily mediated through bioreduction and most of the removed Te was retained in the bioreactor. Characterization using XRD, Raman spectroscopy, SEM-EDX and TEM confirmed association of tellurium with the granular sludge, typically in the form of elemental Te(0) deposits. Furthermore, application of an extracellular polymeric substances (EPS) extraction method to the tellurite reducing sludge recovered up to 78% of the tellurium retained in the granular sludge. This study demonstrates for the first time the application of a UASB reactor for continuous tellurite removal from tellurite-containing wastewater coupled to elemental Te(0) recovery.

  13. Continuous removal and recovery of tellurium in an upflow anaerobic granular sludge bed reactor

    International Nuclear Information System (INIS)

    Mal, Joyabrata; Nancharaiah, Yarlagadda V.; Maheshwari, Neeraj; Hullebusch, Eric D. van; Lens, Piet N.L.

    2017-01-01

    Highlights: • Tellurite bioreduction coupled to recovery of biogenic Te(0) nanocrystals. • First report on continuous tellurite removal in a UASB reactor. • Biogenic Te(0) was mainly associated with loosely-bound EPS of granular sludge. • Repeated exposure to tellurite caused compositional changes in the EPS matrix. - Abstract: Continuous removal of tellurite (TeO 3 2− ) from synthetic wastewater and subsequent recovery in the form of elemental tellurium was studied in an upflow anaerobic granular sludge bed (UASB) reactor operated at 30 °C. The UASB reactor was inoculated with anaerobic granular sludge and fed with lactate as carbon source and electron donor at an organic loading rate of 0.6 g COD L −1 d −1 . After establishing efficient and stable COD removal, the reactor was fed with 10 mg TeO 3 2− L −1 for 42 d before increasing the influent concentration to 20 mg TeO 3 2− L −1 . Tellurite removal (98 and 92%, respectively, from 10 and 20 mg Te L −1 ) was primarily mediated through bioreduction and most of the removed Te was retained in the bioreactor. Characterization using XRD, Raman spectroscopy, SEM-EDX and TEM confirmed association of tellurium with the granular sludge, typically in the form of elemental Te(0) deposits. Furthermore, application of an extracellular polymeric substances (EPS) extraction method to the tellurite reducing sludge recovered up to 78% of the tellurium retained in the granular sludge. This study demonstrates for the first time the application of a UASB reactor for continuous tellurite removal from tellurite-containing wastewater coupled to elemental Te(0) recovery.

  14. Nuclear heat source design for an advanced HTGR process heat plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; O'Hanlon, T.W.

    1983-01-01

    A high-temperature gas-cooled reactor (HTGR) coupled with a chemical process facility could produce synthetic fuels (i.e., oil, gasoline, aviation fuel, methanol, hydrogen, etc.) in the long term using low-grade carbon sources (e.g., coal, oil shale, etc.). The ultimate high-temperature capability of an advanced HTGR variant is being studied for nuclear process heat. This paper discusses a process heat plant with a 2240-MW(t) nuclear heat source, a reactor outlet temperature of 950 0 C, and a direct reforming process. The nuclear heat source outputs principally hydrogen-rich synthesis gas that can be used as a feedstock for synthetic fuel production. This paper emphasizes the design of the nuclear heat source and discusses the major components and a deployment strategy to realize an advanced HTGR process heat plant concept

  15. Molten salt reactors: A new beginning for an old idea

    International Nuclear Information System (INIS)

    LeBlanc, David

    2010-01-01

    Molten salt reactors have seen a marked resurgence of interest over the past decade, highlighted by their inclusion as one of six Generation IV reactor types. The most active development period however was between the mid 1950s and early 1970s at Oak Ridge National Laboratories (ORNL) and any new re-examination of this concept must bear in mind the far different priorities then in place. High breeding ratios and short doubling times were paramount and this guided the evolution of the Molten Salt Breeder Reactor (MSBR) program. As the inherent advantages of the molten salt concept have become apparent to an increasing number of researchers worldwide it is important to not simply look to continue where ORNL left off but to return to basics in order to offer the best design using updated goals and abilities. A major potential change to the traditional Single Fluid, MSBR design and a subject of this presentation is a return to the mode of operation that ORNL proposed for the majority of its MSR program. That being the Two Fluid design in which separate salts are used for fissile 233 UF 4 and fertile ThF 4 . Oak Ridge abandoned this promising route due to what was known as the 'plumbing problem'. It will be shown that a simple yet crucial modification to core geometry can solve this problem and enable the many advantages of the Two Fluid design. In addition, another very promising route laid out by ORNL was simplified Single Fluid converter reactors that could obtain far superior lifetime uranium utilization than LWR or CANDU without the need for any fuel processing beyond simple chemistry control. Updates and potential improvements to this very attractive concept will also be explored.

  16. Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.; Xiang, J.Y.

    1995-01-01

    Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment

  17. Study on severe accident fuel dispersion behavior in the advanced neutron source reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. [Oak Ridge National Lab., TN (United States)] [and others

    1995-09-01

    Core flow blockage events have been determined to represent a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core heatup and melting under full coolant flow condition in other coolant channels. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Hat transfer between melt particle and coolant, which affects the particle breakup characteristics, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to the relative motion of the particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. The results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also they are entrained and move together in a cloud.

  18. Nuclear reactor power for an electrically powered orbital transfer vehicle

    Science.gov (United States)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Kia, T.; Nesmith, B.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low earth orbit (LEO) and geosynchronous earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  19. Nuclear reactor power for an electrically powered orbital transfer vehicle

    International Nuclear Information System (INIS)

    Jaffe, L.; Beatty, R.; Bhandari, P.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low Earth orbit (LEO) and geosynchronous Earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to Earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant

  20. A location procedure for sound sources in reactor-technical enclosures

    International Nuclear Information System (INIS)

    Hamann, D.

    1982-07-01

    A passive method requiring one detector only has been developed for the location of sound emitting faults in nuclear power plant components. It is adapted for use in a frequency range the wavelength of which is of the same order of magnitude as characteristic dimensions of the considered enclosure. The location is performed in the following way: (1) For a fixed detector position the Auto Power Spectral Density (APSD) of the source to be located is measured. (2) For this detector position the APSD is calculated for the potential source locations. For this, the free-field APSD as well as the acoustic normal modes of the enclosure are necessary. (3) The measured APSD is compared with the theoretically obtained APSD's. (4) That APSD is determined which is most similar to the measured APSD, and consequently an information about the unknown source position is got. (author)

  1. Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-01-01

    Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well

  2. The research reactor TRIGA Mainz. A neutron source for versatile applications in research and education

    International Nuclear Information System (INIS)

    Eberhardt, K.; Kronenberg, A.

    2000-01-01

    Currently, four research reactors with a thermal power ranging from 0.1 to 23 MW th are in operation in Germany and one new reactor (20 MW th ) is under construction. The TRIGA Mark II reactor at the Institut fuer Kernchemie became first critical on August 3, 1965. It can be operated in the steady state mode with a maximum power of 100 kW th and in the pulse mode with a peak power of 250 MW th . A survey of the research programmes carried out at the TRIGA Mainz is given covering a wide range of applications in basic and applied science in nuclear chemistry, nuclear- and particle physics. Furthermore, the reactor is used for neutron activation analysis and for education and training of students and technical personal. (orig.) [de

  3. Calculation of the source term for a S1B-sequence at a VVER-1000 type reactor. Part 1

    International Nuclear Information System (INIS)

    Sdouz, G.

    1990-10-01

    The behaviour of the source term in a VVER-1000 type reactor is calculated using the 'Source Term Code Package' (STCP). The input data are based on the russian plant Zaporozhye-5. The selected accident sequence is a small break LOCA in the hot leg followed by loss offsite and onsite electric power (S 1 B-sequence). According to the course of the calculation the results are presented and analyzed for each program. Except for the noble gases all release fractions are lower than 10 -4 . 18 refs., 10 tabs. (Author)

  4. Optimization of binary breeder reactor VI - An acceptable project of binary breeder reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1986-05-01

    A binary breeder reactor that achieves desired characteristics reasonably well has been developed. Its design and characteristics are reported. Previous models showed several complications that result from introduction of 233 U/Th fuel in the core of a LMFBR, compared to purely Pu/U fueled ones. In this new model, the core is made larger to achieve higher breeding ratios and longer refueling intervals, the number of fuel assemblies is increased to accomodate a larger number of control rod assemblies required to compensate for reactivity losses and to control oscillations of the power densities, and, consequently, the fuel inventories are higher. High fuel burnups are achieved without too much complications in the refueling schedule and the power densities can be maintained reasonably constant over an operational cycle. Low sodium void reactivity reduce the potential for severe accidents and a reasonably efficient utilization of thorium can be realized. (Author) [pt

  5. The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1977-01-01

    Concerning refining, fabrication and reprocessing operations of such materials as well as the installation and operation of reactors, necessary regulations are carried out. Namely, in case of establishing the business of refining, fabricating and reprocessing nuclear materials as well as installing nuclear reactors, applications for the permission of the Prime Minister and the Minister of International Trade and Industry should be filed. Change of such operations should be permitted after filing applications. These permissions are retractable. As regards the reactors installed aboard foreign ships, it must be reported to enter Japanese waters and the permission by the Prime Minister must be obtained. In case of nuclear fuel fabricators, a chief technician of nuclear fuel materials (qualified) must be appointed per each fabricator. In case of installing nuclear reactors, the design and methods of construction should be permitted by the Prime Minister. The standard for such permission is specified, and a chief engineer for operating reactors (qualified) must be appointed. Successors inherit the positions of ones who have operated nuclear material refining, fabrication and reprocessing businesses or operated nuclear reactors. (Rikitake, Y.)

  6. Radionuclide inventory and source terms for the surplus production reactors at Hanford

    International Nuclear Information System (INIS)

    Miller, R.L.; Steffes, J.M.

    1987-01-01

    Radionuclide inventories have been estimated for the eight surplus production reactors at Hanford. The inventories listed represent more than 95% of the total curie burden; the remaining 5% is distributed in piping, tunnels, and various other locations within the reactor building and unaccounted for inventories within the reactors or fuel storage basins. Estimates are conservative as the methodology was designed to overestimate the radionuclide inventories in the facilities. The estimated inventory per reactor facility ranges from 13,000 curies to 58,000 curies. The majority of the present inventory consists of tritium, carbon-14, cobalt-60, and nickel-63. The information in this document combines data from past characterization efforts and introduces adjustments for added information and refinement. The inventory of hazardous materials in the reactor facilities is also addressed. This document has been revised to include new reduced inventory figures for chlorine-36. The new figures were derived from recent analysis of irradiated graphite from the 105-kW reactor

  7. Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor; Diseno y construccion de un sistema electronico automatico de medicion y graficado del flujo neutronico para el reactor subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1997-07-01

    The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

  8. An Evolving Worldview: Making Open Source Easy

    Science.gov (United States)

    Rice, Z.

    2017-12-01

    NASA Worldview is an interactive interface for browsing full-resolution, global satellite imagery. Worldview supports an open data policy so that academia, private industries and the general public can use NASA's satellite data to address Earth science related issues. Worldview was open sourced in 2014. By shifting to an open source approach, the Worldview application has evolved to better serve end-users. Project developers are able to have discussions with end-users and community developers to understand issues and develop new features. Community developers are able to track upcoming features, collaborate on them and make their own contributions. Developers who discover issues are able to address those issues and submit a fix. This reduces the time it takes for a project developer to reproduce an issue or develop a new feature. Getting new developers to contribute to the project has been one of the most important and difficult aspects of open sourcing Worldview. After witnessing potential outside contributors struggle, a focus has been made on making the installation of Worldview simple to reduce the initial learning curve and make contributing code easy. One way we have addressed this is through a simplified setup process. Our setup documentation includes a set of prerequisites and a set of straightforward commands to clone, configure, install and run. This presentation will emphasize our focus to simplify and standardize Worldview's open source code so that more people are able to contribute. The more people who contribute, the better the application will become over time.

  9. Conservation as an alternative energy source

    Science.gov (United States)

    Allen, D. E.

    1978-01-01

    A speech is given outlining the energy situation in the United States. It is warned that the existing energy situation cannot prevail and the time is fast running out for continued growth or even maintenance of present levels. Energy conservation measures are given as an aid to decrease U.S. energy consumption, which would allow more time to develop alternative sources of energy.

  10. Plans for an Ultra Cold Neutron source at Los Alamos

    Energy Technology Data Exchange (ETDEWEB)

    Seestrom, S.J.; Bowles, T.J.; Hill, R.; Greene, G.L. [Los Alamos National Lab., NM (United States)

    1996-08-01

    Ultra Cold Neutrons (UCN) can be produced at spallation sources using a variety of techniques. To date the technique used has been to Bragg scatter and Doppler shift cold neutrons into UCN from a moving crystal. This is particularly applicable to short-pulse spallation sources. We are presently constructing a UCN source at LANSCE using method. In addition, large gains in UCN density should be possible using cryogenic UCN sources. Research is under way at Gatchina to demonstrate technical feasibility of be a frozen deuterium source. If successful, a source of this type could be implemented at future spallation source, such as the long pulse source being planned at Los Alamos, with a UCN density that may be two orders of magnitude higher than that presently available at reactors. (author)

  11. DEIMOS – an Open Source Image Database

    Directory of Open Access Journals (Sweden)

    M. Blazek

    2011-12-01

    Full Text Available The DEIMOS (DatabasE of Images: Open Source is created as an open-source database of images and videos for testing, verification and comparing of various image and/or video processing techniques such as enhancing, compression and reconstruction. The main advantage of DEIMOS is its orientation to various application fields – multimedia, television, security, assistive technology, biomedicine, astronomy etc. The DEIMOS is/will be created gradually step-by-step based upon the contributions of team members. The paper is describing basic parameters of DEIMOS database including application examples.

  12. Research reactor safety - an overview of crucial aspects

    International Nuclear Information System (INIS)

    Laverie, M.

    1998-01-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  13. Research reactor safety - an overview of crucial aspects

    Energy Technology Data Exchange (ETDEWEB)

    Laverie, M. [Atomic Energy Commission, Saclay, F-91191 Gif sur Yvette (France)

    1998-07-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  14. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-09-01

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

  15. Random matrix theory with an external source

    CERN Document Server

    Brézin, Edouard

    2016-01-01

    This is a first book to show that the theory of the Gaussian random matrix is essential to understand the universal correlations with random fluctuations and to demonstrate that it is useful to evaluate topological universal quantities. We consider Gaussian random matrix models in the presence of a deterministic matrix source. In such models the correlation functions are known exactly for an arbitrary source and for any size of the matrices. The freedom given by the external source allows for various tunings to different classes of universality. The main interest is to use this freedom to compute various topological invariants for surfaces such as the intersection numbers for curves drawn on a surface of given genus with marked points, Euler characteristics, and the Gromov–Witten invariants. A remarkable duality for the average of characteristic polynomials is essential for obtaining such topological invariants. The analysis is extended to nonorientable surfaces and to surfaces with boundaries.

  16. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Nascimento, Jamil Alves do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  17. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, A. dos; Nascimento, J.A. do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  18. Pseudo-harmonics method: an application to thermal reactors

    International Nuclear Information System (INIS)

    Silva, F.C. da; Rotenberg, S.; Thome Filho, Z.D.

    1985-10-01

    Several applications of the Pseudo-Harmonics method are presented, aiming to calculate the neutron flux and the perturbed eigenvalue of a nuclear reactor, like PWR, with three enrichment regions as Angra-1 reactor. In the reference reactor, perturbations of several types as global as local were simulated. The results were compared with those from the direct calculation. (E.G.) [pt

  19. Design parameters for sludge reduction in an aquatic worm reactor

    NARCIS (Netherlands)

    Hendrickx, T.L.G.; Temmink, B.G.; Elissen, H.J.H.; Buisman, C.J.N.

    2010-01-01

    Reduction and compaction of biological waste sludge from waste water treatment plants (WWTPs) can be achieved with the aquatic worm Lumbriculus variegatus. In our reactor concept for a worm reactor, the worms are immobilised in a carrier material. The size of a worm reactor will therefore mainly be

  20. Saturable reactor-controlled power supply system for TCT/TFTR neutral beam sources

    International Nuclear Information System (INIS)

    Baker, W.R.; Hopkins, D.B.; Dexter, W.L.; Kuenning, R.W.; Smith, B.J.

    1975-11-01

    Each neutral beam source requires one major power supply, the acceleration supply, and four auxiliary power supplies. The power supplies are designed to permit independent interruption of current to any source and crowbarring within 20 μsec, in the event of a source spark, while not disturbing the normal pulsing of all other adjacent sources. The sources are described

  1. Summary of dynamic analyses of the advanced neutron source reactor inner control rods

    International Nuclear Information System (INIS)

    Hendrich, W.R.

    1995-08-01

    A summary of the structural dynamic analyses that were instrumental in providing design guidance to the Advanced Neutron source (ANS) inner control element system is presented in this report. The structural analyses and the functional constraints that required certain performance parameters were combined to shape and guide the design effort toward a prediction of successful and reliable control and scram operation to be provided by these inner control rods

  2. Development of an application simulating radioactive sources

    International Nuclear Information System (INIS)

    Riffault, V.; Locoge, N.; Leblanc, E.; Vermeulen, M.

    2011-01-01

    This paper presents an application simulating radioactive gamma sources developed in the 'Ecole des Mines' of Douai (France). It generates raw counting data as an XML file which can then be statistically exploited to illustrate the various concepts of radioactivity (exponential decay law, isotropy of the radiation, attenuation of radiation in matter). The application, with a spread sheet for data analysis and lab procedures, has been released under free license. (authors)

  3. Forest biomass as an energy source

    Science.gov (United States)

    P.E. Laks; R.W. Hemingway; A. Conner

    1979-01-01

    The Task Force on Forest Biomass as an Energy Source was chartered by the Society of American Foresters on September 26, 1977, and took its present form following an amendment to the charter on October 5, 1977. It built upon the findings of two previous task forces, the Task Force on Energy and Forest Resources and the Task Force for Evaluation of the CORRIM Report (...

  4. Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

    International Nuclear Information System (INIS)

    1978-01-01

    Under the above mentioned law this order prescribes the procedures of controls given to the persons who wish to conduct refining and fabricating businesses, to construct and operate reactors, and to use nuclear source materials, nuclear fuel materials and internationally controlled materials. The common controlling principle prescribed is that the permission or authorization necessary for above listed businesses should be applied for at each factory or each place of business. Based on the principle, the order prescribes: the procedures to apply for the authorization of the refining business, the permission of the change thereof, and the permission of the fabricating business and the change, thereof (the 1st chapter); the procedures to apply for the permission of the construction of reactors and of the change of the construction, as well as the procedure to do periodic inspections of reactor facilities (the 2nd chapter); the procedures to apply for the permission to use nuclear fuel materials and to change the use thereof, the submission of the report to use nuclear source materials, as well as the procedure to apply for the permission to use internationally controlled materials. In the 4th chapter the order lists up the items on which the competent Ministers may require reports from the person who carries on the relevant business. (Matsushima, A.)

  5. 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).

  6. An Overview of Safety Barrier at EPR Reactor in France

    International Nuclear Information System (INIS)

    Mohd Zaid Mohamed; Syahirah Abdul Rahman; Mohamad Azman Che Mat Isa

    2011-01-01

    The fission of atomic nuclei, performed in reactors to generate heat, brings into play large quantities of radiation-emitting radioactive substances from which people and the environment must be protected. This paper explains the need for nuclear safety, which consists of the set of technical and organizational provisions taken at each stage in the design, construction and operation of a nuclear plant to ensure normal service, prevent the risks of an accident and limit its consequences in the unlikely event of its occurrence. (author)

  7. The benefits of an advanced fast reactor fuel cycle for plutonium management

    International Nuclear Information System (INIS)

    Hannum, W.H.; McFarlane, H.F.; Wade, D.C.; Hill, R.N.

    1996-01-01

    The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium and long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a 'focus area' for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed

  8. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  9. Estimating the four-factor product (ε p Pfnl Ptnl) for the accurate calculation of xenon and samarium reactivities in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.

    2007-01-01

    The modified 135 Xe equilibrium reactivity in the Syrian Miniature Neutron Source Reactor (MNSR) was calculated first by using the WIMSD4 and CITATION codes to estimate the four-factor product (ε p P f nl P t nl). Then, precise calculations of 135 Xe and 149 Sm concentrations and reactivities were carried out and compared during the reactor operation time and after shutdown. It was found that the 135 Xe and 149 Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The 149 Sm reactivities could be neglected compared to 135 Xe reactivities during the reactor operating time and after shutdown. (author)

  10. Estimating the four-factor product (ε p Pfnl Ptnl) for the accurate calculation of xenon and samarium reactivities in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.

    2007-01-01

    The modified 135 Xe equilibrium reactivity in the Syrian Miniature Neutron Source Reactor (MNSR) was calculated first by using the WIMSD4 and CITATION codes to estimate the four-factor product (ε p P fnl P tnl ). Then, precise calculations of 135 Xe and 149 Sm concentrations and reactivities were carried out and compared during the reactor operation time and after shutdown. It was found that the 135 Xe and 149 Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The 149 Sm reactivities could be neglected compared to 135 Xe reactivities during the reactor operating time and after shutdown. (author)

  11. Potential of small nuclear reactors for future clean and safe energy sources

    International Nuclear Information System (INIS)

    Sekimoto, H.

    1992-01-01

    To cope with the various kinds of energy demands expected in the 21st century, it is necessary to explore the potential of small nuclear reactors and to find a way of promoting their introduction to society. The main goal of current research activities is 'the constitution of the self-consistent nuclear energy system'. These activities can be understood by realizing that the nuclear community is facing a turning point for its survival in the 21st century. Self-consistency can be manifested by investigating and developing the potential advantages of the nuclear fission reaction and lessening the potential disadvantages. The contributions in this volume discuss concepts of small reactors, applications of small reactors, and consistency with conventional energy supply systems

  12. The contribution of a small TRIGA university research reactor to nuclear research on an international level

    International Nuclear Information System (INIS)

    Villa, M.; Bastuerk, M.; Boeck, H.

    2002-01-01

    The paper focuses especially on the important results in neutron- and solid state physics and the co-operation between the low power TRIGA reactor with high flux neutron sources in Europe such as the Institute Laue-Langevin (ILL) in Grenoble, the Paul Scherrer Institut (PSI) in Villigen, the Rutherford Appleton Laboratory (RAL) in Didcot and the Research Center Juelich. Experiments are set up for test purposes at the TRIGA reactor and then transferred to the powerful neutron sources. Different new perfect silicon channel-cut and interferometer crystals are prepared and then tested at the Bonse-Hart camera, which is a double crystal (or triple axis) diffractometer and at the interferometer set-up. Historically, the first verification of neutron interferometry at a perfect crystal device has been achieved at the 250 kW TRIGA-reactor in Vienna in the year 1974. Also the co-operation with the PSI and the TU Munich in the field of neutron radiography and neutron tomography and VESTA, an experiment for storing cold neutrons with a wavelength of 6.27A, installed at the pulsed neutron source ISIS at RAL are mentioned. The second topic in this paper focuses on the co-operation in the field of safeguard. Several projects have been carried out during the past years in co-operation with the IAEA such as establishing a gamma spectrum reference catalogue for CdZnTe detectors and tests of safeguard video cameras under neutron irradiation. Further an integrated safeguard surveillance network composed of a video camera, a gamma monitor and a neutron monitor is under development

  13. The contribution of a small triga university research reactor to nuclear research on an international level

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Weber, H.W.

    2001-01-01

    The paper focuses especially on the important results in neutron- and solid state physics and the co-operation between the low power TRIGA reactor with high flux neutron sources in Europe such as the Institute Laue-Langevin (ILL) in Grenoble, the Paul Scherrer Institut (PSI) in Villigen, the Rutherford Appleton Laboratory (RAL) in Didcot and the Research Center Juelich. Experiments are set up for test purposes at the TRIGA reactor and then transferred to the powerful neutron sources. Different new perfect silicon channel-cut and interferometer crystals are prepared and then tested at the Bonse-Hart camera, which is a double crystal (or triple axis) diffractometer and at the interferometer set-up. Historically, the first verification of neutron interferometry at a perfect crystal device has been achieved at the 250 kW TRIGA-reactor in Vienna in the year 1974. Also the co-operation with the PSI and the TU Munich in the field of neutron radiography and neutron tomography and VESTA, an experiment for storing cold neutrons with a wavelength of 6.27 A, installed at the pulsed neutron source ISIS at RAL will be mentioned. The second topic treated in this paper shows the international co-operation in the field of superconductors. This research work is carried out under two European TMR-Network programs. The third topic in this paper focuses on the co-operation in the field of safeguard. Several projects have been carried out during the past years in co-operation with the IAEA such as establishing a gamma spectrum reference catalogue for CdZnTe detectors and tests of safeguard video cameras under neutron irradiation. Further an integrated safeguard surveillance network composed of a video camera, a gamma monitor and a neutron monitor is under development. (orig.)

  14. Rotating-bed reactor as a power source for EM gun applications

    Energy Technology Data Exchange (ETDEWEB)

    Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

    1980-01-01

    Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m/sup 3/), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300/sup 0/K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications.

  15. Rotating-bed reactor as a power source for EM gun applications

    International Nuclear Information System (INIS)

    Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

    1980-01-01

    Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m 3 ), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300 0 K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications

  16. Influence of carbon source and inoculum type on anaerobic biomass adhesion on polyurethane foam in reactors fed with acid mine drainage.

    Science.gov (United States)

    Rodriguez, Renata P; Zaiat, Marcelo

    2011-04-01

    This paper analyzes the influence of carbon source and inoculum origin on the dynamics of biomass adhesion to an inert support in anaerobic reactors fed with acid mine drainage. Formic acid, lactic acid and ethanol were used as carbon sources. Two different inocula were evaluated: one taken from an UASB reactor and other from the sediment of a uranium mine. The values of average colonization rates and the maximum biomass concentration (C(max)) were inversely proportional to the number of carbon atoms in each substrate. The highest C(max) value (0.35 g TVS g(-1) foam) was observed with formic acid and anaerobic sludge as inoculum. Maximum colonization rates (v(max)) were strongly influenced by the type of inoculum when ethanol and lactic acid were used. For both carbon sources, the use of mine sediment as inoculum resulted in a v(max) of 0.013 g TVS g(-1) foam day(-1), whereas 0.024 g TVS g(-1) foam day(-1) was achieved with anaerobic sludge. Copyright © 2011 Elsevier Ltd. All rights reserved.

  17. CFD Simulation of an Anaerobic Membrane BioReactor (AnMBR to Treat Industrial Wastewater

    Directory of Open Access Journals (Sweden)

    Laura C. Zuluaga

    2015-06-01

    Full Text Available A Computational Fluid Dynamics (CFD simulation has been developed for an Anaerobic Membrane BioReactor (AnMBR to treat industrial wastewater. As the process consists of a side-stream MBR, two separate simulations were created: (i reactor and (ii membrane. Different cases were conducted for each one, so the surrounding temperature and the total suspended solids (TSS concentration were checked. For the reactor, the most important aspects to consider were the dead zones and the mixing, whereas for the ceramic membrane, it was the shear stress over the membrane surface. Results show that the reactor's mixing process was adequate and that the membrane presented higher shear stress in the 'triangular' channel.

  18. A completely automatic operation type super-safe fast reactor, RAPID. Its application to dispersion source on lunar and earth surfaces

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Kawasaki, Akira; Iwamura, Takamichi

    2002-01-01

    At a viewpoint of flexible measures to future electric power demands, expectation onto a small-scale reactor for dispersion source is increasing gradually. This is thought to increase its importance not only for a source at proximity of its market in advanced nations but also for the one in developing nations. A study on development of the completely automatic operation type super-safe fast reactor, RAPID (refueling by all pins integrated design) has been carried out as a part of the nuclear energy basic research promoting system under three years project since 1999 by a trust of the Japan Atomic Energy Research Institute to a group of the Central Research Institute of Electric Power Industry (CRIEPI) and so on. As the reactor is a lithium cooled fast reactor with 200 Kw of electric output supposing to use at lunar surface, it can be applied to a super-small scale nuclear reactor on the earth, and has feasibility to become a new option of future nuclear power generation. On the other hand, CRIEPI has investigated on various types of fast reactors (RAPID series) for fast reactor for dispersion source on the earth. Here was introduced on such super-safe fast reactors at a center of RAPID-L. (G.K.)

  19. Acoustic source for generating an acoustic beam

    Science.gov (United States)

    Vu, Cung Khac; Sinha, Dipen N.; Pantea, Cristian

    2016-05-31

    An acoustic source for generating an acoustic beam includes a housing; a plurality of spaced apart piezo-electric layers disposed within the housing; and a non-linear medium filling between the plurality of layers. Each of the plurality of piezoelectric layers is configured to generate an acoustic wave. The non-linear medium and the plurality of piezo-electric material layers have a matching impedance so as to enhance a transmission of the acoustic wave generated by each of plurality of layers through the remaining plurality of layers.

  20. End point control of an actinide precipitation reactor

    International Nuclear Information System (INIS)

    Muske, K.R.

    1997-01-01

    The actinide precipitation reactors in the nuclear materials processing facility at Los Alamos National Laboratory are used to remove actinides and other heavy metals from the effluent streams generated during the purification of plutonium. These effluent streams consist of hydrochloric acid solutions, ranging from one to five molar in concentration, in which actinides and other metals are dissolved. The actinides present are plutonium and americium. Typical actinide loadings range from one to five grams per liter. The most prevalent heavy metals are iron, chromium, and nickel that are due to stainless steel. Removal of these metals from solution is accomplished by hydroxide precipitation during the neutralization of the effluent. An end point control algorithm for the semi-batch actinide precipitation reactors at Los Alamos National Laboratory is described. The algorithm is based on an equilibrium solubility model of the chemical species in solution. This model is used to predict the amount of base hydroxide necessary to reach the end point of the actinide precipitation reaction. The model parameters are updated by on-line pH measurements

  1. Treatment of winery wastewater by an anaerobic sequencing batch reactor.

    Science.gov (United States)

    Ruíz, C; Torrijos, M; Sousbie, P; Lebrato Martínez, J; Moletta, R; Delgenès, J P

    2002-01-01

    Treatment of winery wastewater was investigated using an anaerobic sequencing batch reactor (ASBR). Biogas production rate was monitored and permitted the automation of the bioreactor by a simple control system. The reactor was operated at an organic loading rate (ORL) around 8.6 gCOD/L.d with soluble chemical oxygen demand (COD) removal efficiency greater than 98%, hydraulic retention time (HRT) of 2.2 d and a specific organic loading rate (SOLR) of 0.96 gCOD/gVSS.d. The kinetics of COD and VFA removal were investigated for winery wastewater and for simple compounds such as ethanol, which is a major component of winery effluent, and acetate, which is the main volatile fatty acid (VFA) produced. The comparison of the profiles obtained with the 3 substrates shows that, overall, the acidification of the organic matter and the methanisation of the VFA follow zero order reactions, in the operating conditions of our study. The effect on the gas production rate resulted in two level periods separated by a sharp break when the acidification stage was finished and only the breaking down of the VFA continued.

  2. Multivariable robust control of an integrated nuclear power reactor

    Directory of Open Access Journals (Sweden)

    A. Etchepareborda

    2002-12-01

    Full Text Available The design of the main control system of the CAREM nuclear power plant is presented. This plant is an inherently safe low-power nuclear reactor with natural convection on the primary coolant circuit and is self-pressurized with a steam dome on the top of the pressure vessel (PV. It is an integrated reactor as the whole primary coolant circuit is within the PV. The primary circuit transports the heat to the secondary circuit through once-through steam generators (SG. There is a feedwater valve at the inlet of the SG and a turbine valve at the outlet of the SG. The manipulated variables are the aperture of these valves and the reactivity of the control rods. The control target is to regulate the primary and secondary pressures and to monitor steam flow reference ramps on a range of nominal flow from 100% to 40%. The requirements for the control system are robust stability, low-order simple controllers and transient/permanent error bounding. The controller design is based on a detailed RETRAN plant model, from which linear perturbed open-loop dynamic models at different powers are identified. Two low-order nominal models with their associated uncertainties are chosen for two different power ranges. Robust controllers with acceptable performances are designed for each range. Numerical optimization based on the loop-shaping method is used for the controller design. The designed controllers are implemented in the RETRAN model and tested in simulations achieving successful results.

  3. The steam explosion potential for an unseated SRS reactor septifoil

    International Nuclear Information System (INIS)

    Allison, D.K.; Hyder, M.L.; Yau, W.W.F.; Smith, D.C.

    1992-01-01

    Control rods in the Savannah River Site's K Reactor are contained within housings composed of seven channels (''septifoils''). Each septifoil is suspended from the top of the reactor and is normally seated on an upflow pin that channels coolant to the septifoil. Forced flow to the septifoil would be eliminated in the unlikely event of a septifoil unseated upon installation, i.e., if the septifoil is not aligned with its upflow pin. If this event were not detected, control rod melting and the interaction of molten metal with water might occur. This paper describes a methodology used to address the issue of steam explosions that might arise by this mechanism. The probability of occurrence of a damaging steam explosion given an unseated septifoil was found to be extremely low. The primary reasons are: (1) the high probability that melting will not occur, (2) the possibility of material holdup by contact with the outer septifoil housing, (3) the relative shallowness of the pool 'Of water into which molten material might fall, (4) the probable absence of a trigger, and (5) the relatively large energy release required to damage a nearby fuel assembly. The methodology is based upon the specification of conditions prevailing within the septifoil at the time molten material is expected to contact water, and upon information derived from the available experimental data base, supplemented by recent prototypic experiments

  4. Residual stress analysis by neutron time-of-flight at a reactor source

    International Nuclear Information System (INIS)

    Priesmeyer, H.G.; Schroder, J.

    1990-01-01

    Non-destructive neutron diffractometry for stress analysis will be a powerful experimental tool in material science research performed at the GKSS 5 MW reactor FRG-1. Arguments which show the advantages of the time-of-flight method are given and a suitable high-resolution neutron-efficient type of spectrometer is introduced. First results derived from this method are presented

  5. Integrating an Automatic Judge into an Open Source LMS

    Science.gov (United States)

    Georgouli, Katerina; Guerreiro, Pedro

    2011-01-01

    This paper presents the successful integration of the evaluation engine of Mooshak into the open source learning management system Claroline. Mooshak is an open source online automatic judge that has been used for international and national programming competitions. although it was originally designed for programming competitions, Mooshak has also…

  6. Thorium as an energy source. Opportunities for Norway

    Energy Technology Data Exchange (ETDEWEB)

    2008-01-15

    Final Recommendations of the Thorium Report Committee: 1) No technology should be idolized or demonized. All carbon-dioxide (Co2) emission-free energy production technologies should be considered. The potential contribution of nuclear energy to a sustainable energy future should be recognized. 2) An investigation into the resources in the Fen Complex and other sites in Norway should be performed. It is essential to assess whether thorium in Norwegian rocks can be defined as an economical asset for the benefit of future generations. Furthermore, the application of new technologies for the extraction of thorium from the available mineral sources should be studied. 3) Testing of thorium fuel in the Halden Reactor should be encouraged, taking benefit of the well recognized nuclear fuel competence in Halden. 4) Norway should strengthen its participation in international collaborations by joining the EURATOM fission program and the GIF program on Generation IV reactors suitable for the use of thorium. 5) The development of an Accelerator Driven System (ADS) using thorium is not within the capability of Norway working alone. Joining the European effort in this field should be considered. Norwegian research groups should be encouraged to participate in relevant international projects, although these are currently focused on waste management. 6) Norway should bring its competence in waste management up to an international standard and collaboration with Sweden and Finland could be beneficial. 7) Norway should bring its competence with respect to dose assessment related to the thorium cycle up to an international standard. 8) Since the proliferation resistance of uranium-233 depends on the reactor and reprocessing technologies, this aspect will be of key concern should any thorium reactor be built in Norway. 9) Any new nuclear activities in Norway, e.g. thorium fuel cycles, would need strong international pooling of human resources, and in the case of thorium, a strong long

  7. Thorium as an energy source. Opportunities for Norway

    International Nuclear Information System (INIS)

    2008-01-01

    Final Recommendations of the Thorium Report Committee: 1) No technology should be idolized or demonized. All carbon-dioxide (Co2) emission-free energy production technologies should be considered. The potential contribution of nuclear energy to a sustainable energy future should be recognized. 2) An investigation into the resources in the Fen Complex and other sites in Norway should be performed. It is essential to assess whether thorium in Norwegian rocks can be defined as an economical asset for the benefit of future generations. Furthermore, the application of new technologies for the extraction of thorium from the available mineral sources should be studied. 3) Testing of thorium fuel in the Halden Reactor should be encouraged, taking benefit of the well recognized nuclear fuel competence in Halden. 4) Norway should strengthen its participation in international collaborations by joining the EURATOM fission program and the GIF program on Generation IV reactors suitable for the use of thorium. 5) The development of an Accelerator Driven System (ADS) using thorium is not within the capability of Norway working alone. Joining the European effort in this field should be considered. Norwegian research groups should be encouraged to participate in relevant international projects, although these are currently focused on waste management. 6) Norway should bring its competence in waste management up to an international standard and collaboration with Sweden and Finland could be beneficial. 7) Norway should bring its competence with respect to dose assessment related to the thorium cycle up to an international standard. 8) Since the proliferation resistance of uranium-233 depends on the reactor and reprocessing technologies, this aspect will be of key concern should any thorium reactor be built in Norway. 9) Any new nuclear activities in Norway, e.g. thorium fuel cycles, would need strong international pooling of human resources, and in the case of thorium, a strong long

  8. Calculation of anti-seismic design for Xi'an pulsed reactor

    International Nuclear Information System (INIS)

    Li Shuian

    2002-01-01

    The author describes the reactor safety rule, safety regulation and design code that must be observed to anti-seismic design in Xi'an pulsed reactor. It includes the classification of reactor installation, determination of seismic loads, calculate contents, program, method, results and synthetically evaluation. According to the different anti-seismic structure character of reactor installation, an appropriate method was selected to calculate the seismic response. The results were evaluated synthetically using the design code and design requirement. The evaluate results showed that the anti-seismic design function of reactor installation of Xi'an pules reactor is well, and the structure integrality and normal property of reactor installation can be protect under the designed classification of the earthquake

  9. Ion optics in an ion source system

    Energy Technology Data Exchange (ETDEWEB)

    Abdel-Salam, F W; Moustafa, O A; El-Khabeary, H [Accelerators Dept, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    An analysis of ion beams from an ion source which consisted of a hemispherical anode, a plane earthed cathode, and a focusing electrode has been carried out. The focal properties of such electrode arrangement were studied using axially symmetric fields. Axial and radial electric fields were obtained as functions of the axial distance. It was found that the radial component of the gradient of potential pushes the ions towards the axis, which indicates the convergent action of the system. The effect of voltage variation between the boundary and the focusing electrode on the position of the plasma boundary are given using the experimental data of the ion source characteristics and its geometrical parameters. The advantages of plasma diffusing outside the source through a small aperture were used by applying a potential to the focusing electrode. It was possible to extract a large ion current from the expanded plasma. The system constituted a lens with a focal length of 29.4 mm. 7 figs.

  10. Construction of a photochemical reactor combining a CCD spectrophotometer and a LED radiation source.

    Science.gov (United States)

    Gombár, Melinda; Józsa, Éva; Braun, Mihály; Ősz, Katalin

    2012-10-01

    An inexpensive photoreactor using LED light sources and a fibre-optic CCD spectrophotometer as a detector was built by designing a special cell holder for standard 1.000 cm cuvettes. The use of this device was demonstrated by studying the aqueous photochemical reaction of 2,5-dichloro-1,4-benzoquinone. The developed method combines the highly quantitative data collection of CCD spectrophotometers with the possibility of illuminating the sample independently of the detecting light beam, which is a substantial improvement of the method using diode array spectrophotometers as photoreactors.

  11. Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China); Lee, Hyung-Sool [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West Waterloo, Ontario, Canada N2L 3G1 (Canada); Wang, Ai-Jie, E-mail: waj0578@hit.edu.cn [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. Black-Right-Pointing-Pointer Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). Black-Right-Pointing-Pointer PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. Black-Right-Pointing-Pointer The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 {+-} 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m{sup -3} d{sup -1}) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m{sup -3} d{sup -1} (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

  12. Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

    International Nuclear Information System (INIS)

    Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying; Lee, Hyung-Sool; Wang, Ai-Jie

    2012-01-01

    Highlights: ► A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. ► Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). ► PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. ► The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 ± 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m −3 d −1 ) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m −3 d −1 (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

  13. The reformation of liquid hydrocarbons in an aqueous discharge reactor

    KAUST Repository

    Zhang, Xuming

    2015-04-21

    We present an aqueous discharge reactor for the reformation of liquid hydrocarbons. To increase a dielectric constant of a liquid medium, we added distilled water to iso-octane and n-dodecane. As expected, we found decreased discharge onset voltage and increased discharge power with increased water content. Results using optical emission spectroscopy identified OH radicals and O atoms as the predominant oxidative reactive species with the addition of water. Enriched CH radicals were also visualized, evidencing the existence of cascade carbon-carbon cleavage and dehydrogenation processes in the aqueous discharge. The gaseous product consisted primarily of hydrogen, carbon monoxide, and unsaturated hydrocarbons. The composition of the product was readily adjustable by varying the volume of water added, which demonstrated a significant difference in composition with respect to the tested liquid hydrocarbon. In this study, we found no presence of CO2 emissions or the contamination of the reactor by solid carbon deposition. These findings offer a new approach to the reforming processes of liquid hydrocarbons and provide a novel concept for the design of a practical and compact plasma reformer. © 2015 IOP Publishing Ltd.

  14. The reformation of liquid hydrocarbons in an aqueous discharge reactor

    International Nuclear Information System (INIS)

    Zhang, Xuming; Cha, Min Suk

    2015-01-01

    We present an aqueous discharge reactor for the reformation of liquid hydrocarbons. To increase a dielectric constant of a liquid medium, we added distilled water to iso-octane and n-dodecane. As expected, we found decreased discharge onset voltage and increased discharge power with increased water content. Results using optical emission spectroscopy identified OH radicals and O atoms as the predominant oxidative reactive species with the addition of water. Enriched CH radicals were also visualized, evidencing the existence of cascade carbon–carbon cleavage and dehydrogenation processes in the aqueous discharge. The gaseous product consisted primarily of hydrogen, carbon monoxide, and unsaturated hydrocarbons. The composition of the product was readily adjustable by varying the volume of water added, which demonstrated a significant difference in composition with respect to the tested liquid hydrocarbon. In this study, we found no presence of CO 2 emissions or the contamination of the reactor by solid carbon deposition. These findings offer a new approach to the reforming processes of liquid hydrocarbons and provide a novel concept for the design of a practical and compact plasma reformer. (paper)

  15. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.

    2001-01-01

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  16. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.; Voitsekhovitch, I.

    1999-01-01

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  17. An emergency water injection system (EWIS) for future CANDU reactors

    International Nuclear Information System (INIS)

    Marques, Andre L.F.; Todreas, Neil E.; Driscoll, Michael J.

    2000-01-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m 2 : the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  18. Technical feasibility of an Integral Fast Reactor (IFR) as a future option for fast reactor cycles. Integrate a small metal-fueled fast reactor and pyroprocessing facilities

    International Nuclear Information System (INIS)

    Tanaka, Nobuo

    2017-01-01

    Integral Fast Reactor that integrated fast reactor and pyrorocessing facilities developed by Argonne National Laboratory in the U.S. is an excellent nuclear fuel cycle system for passive safety, nuclear non-proliferation, and reduction in radioactive waste. In addition, this system can be considered as a technology applicable to the treatment of the fuel debris caused by the Fukushima Daiichi Nuclear Power Station accident. This study assessed the time required for debris processing, safety of the facilities, and construction cost when using this technology, and examined technological possibility including future technological issues. In a small metal-fueled reactor, it is important to design the core that achieves both of reduction in combustion reactivity and reduction in coolant reactivity. In system design, calorimetric analysis, structure soundness assessment, seismic feasibility establishment study, etc. are important. Regarding safety, research and testing are necessary on the capabilities of passive reactor shutdown and reactor core cooling as well as measures for avoiding re-criticality, even when emergency stop has failed. In dry reprocessing system, studies on electrolytic reduction and electrolytic refining process for treating the debris with compositions different from those of normal fuel are necessary. (A.O.)

  19. An open-source thermodynamic software library

    DEFF Research Database (Denmark)

    Ritschel, Tobias Kasper Skovborg; Gaspar, Jozsef; Capolei, Andrea

    This is a technical report which accompanies the article ”An open-source thermodynamic software library” which describes an efficient Matlab and C implementation for evaluation of thermodynamic properties. In this technical report we present the model equations, that are also presented in the paper......, together with a full set of first and second order derivatives with respect to temperature and pressure, and in cases where applicable, also with respect to mole numbers. The library is based on parameters and correlations from the DIPPR database and the Peng-Robinson and the Soave-Redlich-Kwong equations...

  20. Behavior of radioactive organic iodide in an atmosphere of High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Saeki, Masakatsu; Nakashima, Mikio; Sagawa, Chiaki; Masaki, Nobuyuki; Hirabayashi, Takakuni; Aratono, Yasuyuki

    1990-06-01

    Formation and decomposition behavior of radioactive organic iodide have been studied in an atmosphere of High Temperature Gas-cooled Reactor (High Temperature Engineering Test Reactor, HTTR). Na 125 I was chosen for radioactive iodine source instead of CsI diffusing from coated fuel particles. Na 125 I adsorbed on graphite was heated in pure He and He containing O 2 or H 2 O atmosphere. The results obtained are as follows. It was proved that organic iodide was formed with organic radicals released from graphite even in He atmosphere. Thus, the interchange rate of inorganic iodide with organic iodide was remarkably decreased with prolonged preheat-treatment period at 1000degC. Organic iodide formed was easily decomposed by its recirculation into hot reaction tube kept at 900degC. When organic iodide was passed through powdered graphite bed, more than 70% was decomposed at 90degC. Oxygen and water vapour intermixed in He suppressed the interchange rate of inorganic iodide with organic iodide. These results suggest that organic iodide rarely exists in the pressure vessel under normal operating condition of HTTR, and, under hypothetical accident condition of HTTR, organic iodide fraction never exceeds the value used for a safety assessment of light water reactor. (author)

  1. Nuclear power in space. Use of reactors and radioactive substances as power sources in satellites and space probes

    International Nuclear Information System (INIS)

    Hoestbaeck, Lars

    2008-11-01

    Today solar panels are the most common technique to supply power to satellites. Solar panels will work as long as the power demand of the satellite is limited and the satellite can be equipped with enough panels, and kept in an orbit that allows enough sunlight to hit the panels. There are various types of space missions that do not fulfil these criteria. With nuclear power these types of missions can be powered regardless of the sunlight and as early as 1961 the first satellite with a nuclear power source was placed in orbit. Out of seventy known space missions that has made use of nuclear power, ten have had some kind of failure. In no case has the failure been associated with the nuclear technology used. This report discusses to what degree satellites with nuclear power are a source for potential radioactive contamination of Swedish territory. It is not a discussion for or against nuclear power in space. Neither is it an assessment of consequences if radioactive material from a satellite would reach the earth's surface. Historically two different kinds of Nuclear Power Sources (NPS) have been used to generate electric power in space. The first is the reactor where the energy is derived from nuclear fission of 235 U and the second is the Radioisotope Thermoelectric Generator (RTG) where electricity is generated from the heat of naturally decaying radionuclides. NPS has historically only been used in space by United States and the Soviet Union (and in one failing operation Russia). Nuclear Power Sources have been used in three types of space objects: satellites, space probes and moon/Mars vehicles. USA has launched one experimental reactor into orbit, all other use of NPS by the USA has been RTG:s. The Soviet Union, in contrast, only launched a few RTG:s but nearly forty reactors. The Soviet use of NPS is less transparent than the use in USA and some data published on Soviet systems are more or less well substantiated assessments. It is likely that also future

  2. Analysis of the propagation of neutrons and gamma-rays from the fast neutron source reactor YAYOI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Shigeo, E-mail: neutron@keyaki.cc.u-tokai.ac.jp [Department of Energy Science and Engineering, School of Engineering, Tokai University, Hiratsuka, Kanagawa 259-1292 (Japan); Murata, Isao [Division of Electrical, Electronic and Information Engineering, Osaka University, Suita, Osaka 565-0871 (Japan); Nakagawa, Tsutomu; Saito, Isao [Nuclear Professional School, School of Engineering, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki 319-1188 (Japan)

    2011-10-01

    The skyshine effect is crucial for designing appropriate shielding. To investigate the skyshine effect, the propagation of neutrons was measured and analyzed at the fast neutron source reactor YAYOI. Pulse height spectra and dose distributions of neutron and secondary gamma-ray were measured outside YAYOI, and analyzed with MCNP-5 and JENDL-3.3. Comparison with the experimental results showed good agreement. Also, a semi-empirical formula was successfully derived to describe the dose distribution. The formulae can be used to predict the skyshine effect at YAYOI, and will be useful for estimating the skyshine effect and designing the shield structure for fusion facilities.

  3. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    International Nuclear Information System (INIS)

    Ruggles, A.E.; Cheng, L.Y.; Dimenna, R.A.; Griffith, P.; Wilson, G.E.

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis

  4. Solid state laser driver for an ICF reactor

    International Nuclear Information System (INIS)

    Krupke, W.F.

    1988-01-01

    A conceptual design is presented of the main power amplifier of a multi-beamline, multi-megawatt solid state ICF reactor driver. Simultaneous achievement of useful beam quality and high average power is achieved by a proper choice of amplifier geometry. An amplifier beamline consists of a sequence of face-pumped rectangular slab gain elements, oriented at the Brewster angle relative to the beamline axis, and cooled on their large faces by helium gas that is flowing subsonically. The infrared amplifier output radiation is shifted to an appropriately short wavelength ( 10% (including all flow cooling input power) when the amplifiers are pumped by efficient high-power AlGaAs semiconductor laser diode arrays. 11 refs., 3 figs., 7 tabs

  5. Feedback of reactor operating data to nuclear methods development

    International Nuclear Information System (INIS)

    Crowther, R.L.; Kang, C.M.; Parkos, G.R.; Wolters, R.A.

    1978-01-01

    The problems in obtaining power reactor data for reliable nuclear methods development and the major sources of power reactor data for this purpose are reviewed. Specific examples of the use of power reactor data in nuclear methods development are discussed. The paper concludes with recommendations on the key elements of an effective program to use power reactor data in nuclear methods development

  6. Comparative study on nutrient removal of agricultural non-point source pollution for three filter media filling schemes in eco-soil reactors.

    Science.gov (United States)

    Du, Fuyi; Xie, Qingjie; Fang, Longxiang; Su, Hang

    2016-08-01

    Nutrients (nitrogen and phosphorus) from agricultural non-point source (NPS) pollution have been increasingly recognized as a major contributor to the deterioration of water quality in recent years. The purpose of this article is to investigate the discrepancies in interception of nutrients in agricultural NPS pollution for eco-soil reactors using different filling schemes. Parallel eco-soil reactors of laboratory scale were created and filled with filter media, such as grit, zeolite, limestone, and gravel. Three filling schemes were adopted: increasing-sized filling (I-filling), decreasing-sized filling (D-filling), and blend-sized filling (B-filling). The systems were intermittent operations via simulated rainstorm runoff. The nutrient removal efficiency, biomass accumulation and vertical dissolved oxygen (DO) distribution were defined to assess the performance of eco-soil. The results showed that B-filling reactor presented an ideal DO for partial nitrification-denitrification across the eco-soil, and B-filling was the most stable in the change of bio-film accumulation trends with depth in the three fillings. Simultaneous and highest removals of NH4(+)-N (57.74-70.52%), total nitrogen (43.69-54.50%), and total phosphorus (42.50-55.00%) were obtained in the B-filling, demonstrating the efficiency of the blend filling schemes of eco-soil for oxygen transfer and biomass accumulation to cope with agricultural NPS pollution.

  7. Closed loop auto control system software for Miniature Neutron Source Reactors (MNSRs)

    International Nuclear Information System (INIS)

    Iqbal, Masood; Qadir, Javed

    2009-01-01

    A closed loop auto control system has been developed and implemented at the Pakistan Research Reactor-2 (PARR-2). For interfacing of the site signals with a computer, a data acquisition card was employed. The system utilizes nine signals and, on their basis, carries out continuous on-line analysis, maintaining safe operation. On-line reactivity measurement is included in this software. Also, it generates control signals for keeping the reactor at the desired power level in auto-mode round-the-clock without human intervention. In case of abnormal conditions for either power or temperature or radiation level, alarms are initiated and if their levels reach beyond prescribed safe operation regime, automatic shutdown sequence is initiated. The control system has been thoroughly tested for various postulated scenarios. The test results have been in good agreement with the expected response. (orig.)

  8. Study of an hypothetical reactor meltdown accident for a 50 MW sub(th) fast reactor

    International Nuclear Information System (INIS)

    Azevedo, E.M. de.

    1983-01-01

    A melhodology for determining the energy released in hypothetical reactor meltdown accidents is presented. A numerical code was developed based upon the Nicholson method for a uniform and homogeneous reactor with spherical geometry. A comparative study with other know programs in the literature which use better approximations for small energy released, shows that the methodology used were compatible with those under comparison. Besides the influence of some parameters on the energy released, such as the initial power level and the prompt neutron lifetime was studied under this metodology and its result exhibitted. The Doppler effect was also analyzed and its influence on the energy released has been emphasized. (Author) [pt

  9. Modeling and analysis of hydrogen detonation events in the advanced neutron source reactor containment

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.; Georgevich, V.; Kim, S.H.; Valenti, S.; Simpson, D.B.; Sawruk, W.

    1994-01-01

    This paper describes salient aspects of the modeling, analyses, and evaluations for hydrogen detonation in selected regions of the Advanced Neutron Source (ANS) containment during hypothetical severe accident conditions. Shock wave generation and transport modeling and analyses were conducted for two stratified configurations in the dome region of the high bay. Principal tools utilized for these purposes were the CTH and CET89 computer codes. Dynamic pressure loading functions were generated for key locations and used for evaluating structural response behavior for which a finite-element model was developed using the ANSYS code. For the range of conditions analyzed in the two critical dome regions, it was revealed that the ANS containment would be able to withstand detonation loads without failure. (author)

  10. Modeling and analysis of hydrogen detonation events in the Advanced Neutron Source reactor containment

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.; Georgevich, V.; Kim, S.H.; Valenti, S.N.; Simpson, D.B.; Sawruk, W.

    1994-07-01

    This paper describes salient aspects of the modeling, analyses, and evaluations for hydrogen detonation in selected regions of the Advanced Neutron Source (ANS) containment during hypothetical severe accident conditions. Shock wave generation and transport modeling and analyses were conducted for two stratified configurations in the dome region of the high bay. Principal tools utilized for these purposes were the CTH and CET89 computer codes. Dynamic pressure loading functions were generated for key locations and used for evaluating structural response behavior for which a finite-element model was developed using the ANSYS code. For the range of conditions analyzed in the two critical dome regions, it was revealed that the ANS containment would be able to withstand detonation loads without failure

  11. Modeling and simulation of an isothermal reactor for methanol steam reforming

    Directory of Open Access Journals (Sweden)

    Raphael Menechini Neto

    2014-04-01

    Full Text Available Due to growing electricity demand, cheap renewable energy sources are needed. Fuel cells are an interesting alternative for generating electricity since they use hydrogen as their main fuel and release only water and heat to the environment. Although fuel cells show great flexibility in size and operating temperature (some models even operate at low temperatures, the technology has the drawback for hydrogen transportation and storage. However, hydrogen may be produced from methanol steam reforming obtained from renewable sources such as biomass. The use of methanol as raw material in hydrogen production process by steam reforming is highly interesting owing to the fact that alcohol has the best hydrogen carbon-1 ratio (4:1 and may be processed at low temperatures and atmospheric pressures. They are features which are desirable for its use in autonomous fuel cells. Current research develops a mathematical model of an isothermal methanol steam reforming reactor and validates it against experimental data from the literature. The mathematical model was solved numerically by MATLAB® and the comparison of its predictions for different experimental conditions indicated that the developed model and the methodology for its numerical solution were adequate. Further, a preliminary analysis was undertaken on methanol steam reforming reactor project for autonomous fuel cell.

  12. Operation and reactivity measurements of an accelerator driven subcritical TRIGA reactor

    Science.gov (United States)

    O'Kelly, David Sean

    Experiments were performed at the Nuclear Engineering Teaching Laboratory (NETL) in 2005 and 2006 in which a 20 MeV linear electron accelerator operating as a photoneutron source was coupled to the TRIGA (Training, Research, Isotope production, General Atomics) Mark II research reactor at the University of Texas at Austin (UT) to simulate the operation and characteristics of a full-scale accelerator driven subcritical system (ADSS). The experimental program provided a relatively low-cost substitute for the higher power and complexity of internationally proposed systems utilizing proton accelerators and spallation neutron sources for an advanced ADSS that may be used for the burning of high-level radioactive waste. Various instrumentation methods that permitted ADSS neutron flux monitoring in high gamma radiation fields were successfully explored and the data was used to evaluate the Stochastic Pulsed Feynman method for reactivity monitoring.

  13. An approach of raising the low power reactor trip block (P-7) in Maanshan Power Plant

    International Nuclear Information System (INIS)

    Wang, L.C.

    1984-01-01

    The technical specification for the Maanshan Nuclear Power Station (FSAR Table 16.2.2-3) requires that with an increasing reactor power level above the setpoint of low power reactor trip block (P-7), a turbine trip shall initiate a reactor trip. This anticipatory reactor trip on turbine trip prevents the pressurizer PORV from openning during turbine trip event. In order to reduce unnecessary reactor trip due to turbine trip on low reactor power level during Maanshan start-up stage, Taiwan Power Company performed a transient analysis for turbine trip event by using RETRAN code. The highest reactor power level at which a turbine trip will not open the pressurizer PORV is searched. The results demonstrated that this power level can be increased from the original value-10% of the rated thermal power-to about 48% of the rated thermal power

  14. IDAS-RR: an incident data base system for research reactors

    International Nuclear Information System (INIS)

    Matsumoto, Kiyoshi; Kohsaka, Atsuo; Kaminaga, Masanori; Murayama, Youji; Ohnishi, Nobuaki; Maniwa, Masaki.

    1990-03-01

    An Incident Data Base System for Research Reactors, IDAS-RR, has been developed. IDAS-RR has information about abnormal incidents (failures, transients, accidents, etc.) of research reactors in the world. Data reference, input, editing and other functions of IDAS-RR are menu driven. The routine processing and data base management functions are performed by the system software and hardware. PC-9801 equipment was selected as the hardware because of its portability and popularity. IDAS-RR provides effective reference information for the following activities. 1) Analysis of abnormal incident of research reactors, 2) Detail analysis of research reactor behavior in the abnormal incident for building the knowledge base of the reactor emergency diagnostic system for research reactor, 3) Planning counter-measure for emergency situation in the research reactor. This report is a user's manual of IDAS-RR. (author)

  15. Research reactors serving the radiotracer technique: An overlooked opportunity?

    International Nuclear Information System (INIS)

    Bode, P.

    2010-01-01

    The unique features of using radiotracers are not always fully explored at nuclear research reactor centers. The radiotracer method is a versatile and powerful tool in the study of a wide variety of applications in e.g. chemistry, biology, agriculture, medicine and (industrial) technology. The big advantage of radiotracers above e.g. stable isotope tracers, commonly applied in e.g. nutritional studies, is that radiotracers allow for non-invasive studies of both steady-state and dynamic systems, in equilibrium situations and for transport and exchange phenomena and thus provide information on the chemical and/or physical status of elements. The radiotracer method does not imply huge equipment investments but rather requires that the four interrelated aspects: experimental designs, data treatment including tracer kinetic analysis and data interpretation are careful considered. Radiotracer production with reactors often implies additional fundamental research to use smart nuclear reactions and chemical separations so as to obtain an almost no-carrier added tracer. Nuclear analytical groups, already equipped with gamma-ray spectrometers, can thus extend their research program. Using Ge-spectrometers, multi-labeling experiments are possible which allow for unique applications if different radionuclides exist for the same element, such as 65 Zn and 69m Zn or 64 Cu and 67 Cu. The developments in novel scintillation detectors, room-temperature semiconductor detectors (like CdZnTe) and position-sensitive detectors open the door for an entire new scope of radiotracer applications, such as SPECT. Once applied to in-vitro studies with cell cultures, the nuclear analytical group may position itself into the worlds of medical research, biochemistry and biotechnology. Radiotracers can be used to study the properties of drug-delivering compounds as used in e.g, cancer therapy. Radiotracers can be added as a label to solid particles, to liquids but may also be applied in the

  16. Modeling Geometric Arrangements of TiO2-Based Catalyst Substrates and Isotropic Light Sources to Enhance the Efficiency of a Photocatalystic Oxidation (PCO) Reactor

    Science.gov (United States)

    Richards, Jeffrey T.; Levine, Lanfang H.; Husk, Geoffrey K.

    2011-01-01

    The closed confined environments of the ISS, as well as in future spacecraft for exploration beyond LEO, provide many challenges to crew health. One such challenge is the availability of a robust, energy efficient, and re-generable air revitalization system that controls trace volatile organic contaminants (VOCs) to levels below a specified spacecraft maximum allowable concentration (SMAC). Photocatalytic oxidation (PCO), which is capable of mineralizing VOCs at room temperature and of accommodating a high volumetric flow, is being evaluated as an alternative trace contaminant control technology. In an architecture of a combined air and water management system, placing a PCO unit before a condensing heat exchanger for humidity control will greatly reduce the organic load into the humidity condensate loop ofthe water processing assembly (WPA) thereby enhancing the life cycle economics ofthe WPA. This targeted application dictates a single pass efficiency of greater than 90% for polar VOCs. Although this target was met in laboratory bench-scaled reactors, no commercial or SBIR-developed prototype PCO units examined to date have achieved this goal. Furthermore, the formation of partial oxidation products (e.g., acetaldehyde) was not eliminated. It is known that single pass efficiency and partial oxidation are strongly dependent upon the contact time and catalyst illumination, hence the requirement for an efficient reactor design. The objective of this study is to maximize the apparent contact time and illuminated catalyst surface area at a given reactor volume and volumetric flow. In this study, a Ti02-based photocatalyst is assumed to be immobilized on porous substrate panels and illumination derived from linear isotropic light sources. Mathematical modeling using computational fluid dynamics (CFD) analyses were performed to investigate the effect of: 1) the geometry and configuration of catalyst-coated substrate panels, 2) porosity of the supporting substrate, and 3

  17. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Trial Calculation. Work Plan

    International Nuclear Information System (INIS)

    Grabaskas, David; Bucknor, Matthew; Jerden, James; Brunett, Acacia J.

    2016-01-01

    The overall objective of the SFR Regulatory Technology Development Plan (RTDP) effort is to identify and address potential impediments to the SFR regulatory licensing process. In FY14, an analysis by Argonne identified the development of an SFR-specific MST methodology as an existing licensing gap with high regulatory importance and a potentially long lead-time to closure. This work was followed by an initial examination of the current state-of-knowledge regarding SFR source term development (ANLART-3), which reported several potential gaps. Among these were the potential inadequacies of current computational tools to properly model and assess the transport and retention of radionuclides during a metal fuel pool-type SFR core damage incident. The objective of the current work is to determine the adequacy of existing computational tools, and the associated knowledge database, for the calculation of an SFR MST. To accomplish this task, a trial MST calculation will be performed using available computational tools to establish their limitations with regard to relevant radionuclide release/retention/transport phenomena. The application of existing modeling tools will provide a definitive test to assess their suitability for an SFR MST calculation, while also identifying potential gaps in the current knowledge base and providing insight into open issues regarding regulatory criteria/requirements. The findings of this analysis will assist in determining future research and development needs.

  18. FAFTRCS: an experiment in computerized reactor safety systems

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1985-01-01

    Nuclear Power Plant availability and reliability could be improved by the integration of computers into the control environment. However, computer-based systems are historically viewed as being unreliable. This places a burden upon the designer to demonstrate adequate reliability and availability for the computer. The complexity associated with computers coupled with the manual nature of these demonstrations results in a high cost which typically has been justified for critical applications only. This paper investigates a methodology for automating this process and discusses a project which intends to apply this methodology to design verification and validation for a control system which will be installed and tested in an actual reactor control environment. 7 refs., 4 figs., 1 tab

  19. An evaluation on environment radiation impact of pulsed reactor

    International Nuclear Information System (INIS)

    Gao Yingwei; Pu Gongxu; Li Jian

    1991-01-01

    The dose regulation, assessment scope and assessment method adopted by the environment impact evaluation for the pulsed reactor are discussed. The compute model, the compute programme and the compute result of the dose adopted for the model pulsed reactor are introduced. The probable environment radiation impact under normal status and accident status are also appraised

  20. An analysis of the falling film gas-liquid reactor

    NARCIS (Netherlands)

    Davis, E.J.; Ouwerkerk-Dijkers, van M.P.; Venkatesh, S.

    1979-01-01

    A mathematical model of the falling film reactor is developed to predict the conversion and temperature distribution in the reactor as a function of the gas and liquid flow rates, physical properties, the feed composition of the reactive gas and carrier gas and other parameters of the system.