WorldWideScience

Sample records for pulsed reactor facilities

  1. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    International Nuclear Information System (INIS)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters

  2. THE WHITE SANDS MISSILE RANGE PULSED REACTOR FACILITY, MAY 1963

    Energy Technology Data Exchange (ETDEWEB)

    Long, Robert L.; Boor, R. A.; Cole, W. M.; Elder, G. E.

    1963-05-15

    A brief statement of the mission of the White Sands Missile Range Nuclear Effects Laboratory is given. The new Nuclear Effects Laboratory Facility is described. This facility consists of two buildings-a laboratory and a reactor building. The White Sands Missile Range bare critical assembly, designated as the MoLLY-G, is described. The MoLLY-G, an unreflected, unmoderated right circular cylinder of uranium-molybdenum alloy designed for pulsed operation, will have a maximum burst capability of approximately 2 x 10/sup 17/ fissions with a burst width of 50 microseconds. The reactor construction and operating procedures are described. As designed, the MoLLY-G will provide an intense source of pulsed neutron and gamma radiation for a great variety of experimental and test arrangements. (auth)

  3. Test and application of thermal neutron radiography facility at Xi'an pulsed reactor

    CERN Document Server

    Yang Jun; Zhao Xiang Feng; Wang Dao Hua

    2002-01-01

    A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter

  4. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  5. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    1975-01-01

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

  6. Pulsed Compression Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roestenberg, T. [University of Twente, Enschede (Netherlands)

    2012-06-07

    The advantages of the Pulsed Compression Reactor (PCR) over the internal combustion engine-type chemical reactors are briefly discussed. Over the last four years a project concerning the fundamentals of the PCR technology has been performed by the University of Twente, Enschede, Netherlands. In order to assess the feasibility of the application of the PCR principle for the conversion methane to syngas, several fundamental questions needed to be answered. Two important questions that relate to the applicability of the PCR for any process are: how large is the heat transfer rate from a rapidly compressed and expanded volume of gas, and how does this heat transfer rate compare to energy contained in the compressed gas? And: can stable operation with a completely free piston as it is intended with the PCR be achieved?.

  7. Physical model of reactor pulse

    International Nuclear Information System (INIS)

    Petrovic, A.; Ravnik, M.

    2004-01-01

    Pulse experiments have been performed at J. Stefan Institute TRIGA reactor since 1991. In total, more than 130 pulses have been performed. Extensive experimental information on the pulse physical characteristics has been accumulated. Fuchs-Hansen adiabatic model has been used for predicting and analysing the pulse parameters. The model is based on point kinetics equation, neglecting the delayed neutrons and assuming constant inserted reactivity in form of step function. Deficiencies of the Fuchs-Hansen model and systematic experimental errors have been observed and analysed. Recently, the pulse model was improved by including the delayed neutrons and time dependence of inserted reactivity. The results explain the observed non-linearity of the pulse energy for high pulses due to finite time of pulse rod withdrawal and the contribution of the delayed neutrons after the prompt part of the pulse. The results of the improved model are in good agreement with experimental results. (author)

  8. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

    2010-01-01

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  9. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  10. PUSPATI Triga Reactor pulsing parameters

    Energy Technology Data Exchange (ETDEWEB)

    Auu, Gui Ah; Abu, Puad Haji; Yunus, Yaziz [PUSPATI, Selangor (Malaysia)

    1984-06-01

    The pulsing experiment was carried out as part of the commissioning activites of PUSPATI TRIGA Reactor (PTR). Several parameters of PTR were deduced from the experiment. It was found that the maximum temperature of the fuel was far below the safety limit when the maximum allowable positive reactivity of $3.00 was inserted into the core. The peak power achieved was 1354 Mw.

  11. Nuclear reactor facility

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    In order to improve the performance of manitenance and inspections it is proposed for a nuclear reactor facility with a primary circuit containing liquid metal to provide a thermally insulated chamber, within which are placed a number of components of the primary circuit, as e.g. valves, recirculation pump, heat exchangers. The isolated placement permit controlled preheating on one hand, but prevents undesirable heating of adjacent load-bearing elements on the other. The chamber is provided with heating devices and, on the outside, with cooling devices; it is of advantage to fill it with an inert gas. (UWI) 891 HP [de

  12. Pulse coil concepts for the LCP Facility

    International Nuclear Information System (INIS)

    Nelson, B.E.; Burn, P.B.

    1977-01-01

    The pulse coils described in this paper are resistive copper magnets driven by time-varying currents. They are included in the Large Coil Test Facility (LCTF) portion of the Large Coil Program (LCP) to simulate the pulsed field environment of the toroidal coils in a tokamak reactor. Since TNS (a 150 sec, 5MA, igniting tokamak) and the Oak Ridge EPR (Experimental Power Reactor) are representative of the first tokamaks to require the technology developed in LCP, the reference designs for these machines, especially TNS, are used to derive the magnetic criteria for the pulse coils. This criteria includes the magnitude, distribution, and rate of change of pulsed fields in the toroidal coil windings. Three pulse coil concepts are evaluated on the basis of magnetic criteria and factors such as versatility of design, ease of fabrication and cost of operation. The three concepts include (1) a pair of poloidal coils outside the LCTF torus, (2) a single poloidal coil threaded through the torus, and (3) a pair of vertical axis coil windings inside the bore of one or more of the toroidal test coils

  13. PUSPATI Triga Reactor pulsing parameters

    International Nuclear Information System (INIS)

    Gui Ah Auu; Puad Haji Abu; Yaziz Yunus

    1984-01-01

    The pulsing experiment was carried out as part of the commissioning activites of PUSPATI TRIGA Reactor (PTR). Several parameters of PTR were deduced from the experiment. It was found that the maximum temperature of the fuel was far below the safety limit when the maximum allowable positive reactivity of $3.00 was inserted into the core. The peak power achieved was 1354 Mw. (author)

  14. Reactor feedwater facility

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Tadashi; Kinoshita, Shoichiro; Akatsu, Jun-ichi

    1996-04-30

    In a reactor feedwater facility in which one stand-by system and at least three ordinary systems are disposed in parallel, each of the feedwater pumps is driven by an electromotor, and has substantially the same capacity. At least two systems among the ordinary systems have a pump rotation number variable means. Since the volume of each of the feedwater pump of each system is determined substantially equal, standardization is enabled to facilitate the production. While the number of electromotors is increased, since they are driven by electromotors, turbines, steam pipelines and valves for driving feed water pumps can be eliminated. Therefore, the feedwater pumps can be disposed to a region of low radiation dose being separated from a main turbine and a main condensator, to improve the degree of freedom in view of the installation. In addition, accessibility to equipments during operation is improved to improve the maintenance of feed water facilities. The number of parts for equipments can be reduced compared with that in a turbine-driving system thereby capable of reducing the operation amount for the maintenance and inspection. (N.H.)

  15. Upgrade of the Annular Core Pulse Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reuscher, J A [Sandia Laboratories, Albuquerque, NM (United States)

    1976-07-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past two years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 by utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. Preliminary studies have identified several potential approaches to the ACPR performance improvement. The most promising approach appears to be the two-region core concept. The inner region, surrounding the irradiation cavity, would consist of a high-heat capacity fuel capable of absorbing the fission energy associated with a large nuclear pulse. The number of fissions occurring near the cavity would be greatly increased which, in turn, would significantly increase the fluence in the cavity. The outer region would consist of a U-ZrH fuel to provide an overall negative temperature coefficient for the two region core. Two candidate high heat capacity fuels [(BeO-UO{sub 2} and UC-ZrC) - graphite] are under consideration. Since this reactor upgrade represents a modification to an existing facility, it can be achieved in a relatively short time. It is anticipated that most of the existing reactor structure can be used for the upgrade. The present core occupies about one-half of the location in the grid plate. The high-heat capacity fuel

  16. The Toulouse pulsed magnet facility

    International Nuclear Information System (INIS)

    2006-01-01

    The 'Laboratoire National des Champs Magnetiques Pulses' (LNCMP) is an international user facility providing access to pulsed magnetic fields up to and beyond 60 T. The laboratory disposes of 10 magnet stations equipped with long-pulse magnets operating in the 35-60 T range and a short-pulse system reaching magnetic fields in excess of 70 T. The experimental infrastructure includes various high and low-temperature systems ranging from ordinary flow-type cryostats to dilution refrigerators reaching 50 mK, as well as different types of high-pressure cells. Experimental techniques include magnetization, transport, luminescence, IR-spectroscopy and polarimetry. The LNCMP pursues an extensive in-house research program focussing on all technological and scientific aspects of pulsed magnetic fields. Recent technical developments include the implementation of 60 T rapid-cooling coils, an 80 T prototype, a pulsed dipole magnet for optical investigations of dilute matter and a transportable horizontal access magnet for small angle x-ray scattering experiments. Scientific activities cover a variety of domains, including correlated electron systems, magnetism, semiconductors and nanoscience

  17. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

    The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

  18. Sandia Pulse Reactor-IV Project

    International Nuclear Information System (INIS)

    Reuscher, J.A.

    1983-01-01

    Sandia National Laboratories has developed, designed and operated fast burst reactors for over 20 years. These reactors have been used for a variety of radiation effects programs. During this period, programs have required larger irradiation volumes primarily to expose complex electronic systems to postulated threat environments. As experiment volumes increased, a new reactor was built so that these components could be tested. The Sandia Pulse Reactor-IV is a logical evolution of the two decades of fast burst reactor development at Sandia

  19. Reactor containing facility

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1992-01-01

    A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

  20. Reactor container facility

    International Nuclear Information System (INIS)

    Saito, Takashi; Nagasaka, Hideo.

    1990-01-01

    A dry-well pool for spontaneously circulating stored pool water and a suppression pool for flooding a pressure vessel by feeding water, when required, to a flooding gap by means of spontaneous falling upto the flooding position, thereby flooding the pressure vessel are contained at the inside of a reactor container. Thus, when loss of coolant accidents such as caused by main pipe rupture accidents should happen, pool water in the suppression pool is supplied to the flooding gap by spontaneously falling. Further, if the flooding water uprises exceeding a predetermined level, the flooding gap is in communication with the dry-well pool at the upper and the lower portions respectively. Accordingly, flooding water at high temperature heated by the after-heat of the reactor core is returned again into the flooding gap to cool the reactor core repeatedly. (T.M.)

  1. Reactor water injection facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro

    1997-05-02

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  2. Emergency reactor core cooling facility

    International Nuclear Information System (INIS)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka.

    1996-01-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  3. Emergency reactor core cooling facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

    1996-11-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  4. Reactor fuel exchanging facility

    International Nuclear Information System (INIS)

    Kubota, Shin-ichi.

    1981-01-01

    Purpose: To enable operation of an emergency manual operating mechanism for a fuel exchanger with all operatorless trucks and remote operation of a manipulator even if the exchanger fails during the fuel exchanging operation. Constitution: When a fuel exchanging system fails while connected to a pressure tube of a nuclear reactor during a fuel exchanging operation, a stand-by self-travelling truck automatically runs along a guide line to the position corresponding to the stopping position at that time of the fuel exchanger based on a command from a central control chamber. At this time the truck is switched to manual operation, and approaches the exchanger while being monitored through a television camera and then stops. Then, a manipurator is connected to the emergency manual operating mechanism of the exchanger, and is operated through necessary emergency steps by driving the snout, the magazine, the grab or the like in the exchanger in response to the problem, and necessary operations for the emergency treatment are thus performed. (Sekiya, K.)

  5. Emergency reactor container cooling facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Matsumoto, Tomoyuki.

    1992-01-01

    The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

  6. Nuclear Reactor Facility

    International Nuclear Information System (INIS)

    Schabert, H.P.; Ropers, J.

    1976-01-01

    A pressurized-water reactor pressure vessel connects via a main coolant pipe loop including a main coolant pump, with the lower portion of at least one vertical steam generator horizontally offset from the pressure vessel. This equipment is contained by a concrete structure entirely enclosing the pressure vessel and forming a generator room horizontally enclosing the generator and the loop and extending upwardly to an open top closed by a horizontal ceiling. The concrete structure is completely surrounded by a spherical steel containment shell designed to withstand any internal fluid pressure which might result from an accidental release of the coolant inside of this shell, and the shell forms a large space above the entire concrete structure. The ceiling above the generator room is a horizontal steel gridlike construction defining a plurality of vertical openings which are normally closed by horizontal sheet metal plates which are hinged to the gridlike construction and are light enough in weight to be forced upwardly, to open the openings, when the plates receive upward force from fluid pressure below them resulting from the loop, or other equipment in the generator room, accidentally permitting a sudden release of the pressurized-water coolant. The high fluid pressure that would otherwise develop within the concrete generator room, is in this way almost immediately relieved via the openings of the grid-like construction, by the plates being forced upwardly, the pressure being then dissipated upwardly in the large space above the top of the concrete structure, provided by the steel containment shell. This prevents the upstanding wall portions of the generator room from being stressed, and possibly damaged, by any sudden release of coolant in the generator room. Other features are disclosed

  7. 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

  8. Pulse radiolysis facilities and activities in Japan

    International Nuclear Information System (INIS)

    Ogasawara, M.

    1995-01-01

    Pulse radiolysis studies in Japan have been reviewed in special reference to the facilities and the people who have engaged in the experiments. Main achievement is summarized with the list of selected publications. (author)

  9. Reactor kinetics - pulse and steady state

    Energy Technology Data Exchange (ETDEWEB)

    Estes, B F; Morris, F M [Sandia Laboratories (United States)

    1974-07-01

    An analytical model has been developed which couples the nuclear and thermal characteristics of the Annular Core Pulse Reactor (ACPR) into a solution which describes both the neutron kinetics of the reactor and the temperature behavior of a fuel-moderator element. The model describes both pulse and steady state operations. This paper describes the important aspects of the reactor, the fuel- moderator elements, the neutron kinetic equations of the reactor, and the time-temperature behavior of a fuel-moderator element that is being subjected to the maximum power density in the core. The parameters which are utilized in the equations are divided into two classes, those that can be measured directly and those that are assumed to be known (each is described briefly). Some of the solutions which demonstrate the versatility of the analytical model are described. (author)

  10. Reactor Sharing at Rensselaer Critical Facility

    International Nuclear Information System (INIS)

    D. Steiner, D. Harris, T. Trumbull

    2006-01-01

    This final report summarizes the reactor sharing activities at the Rensselaer Critical Facility. An example of a typical tour is also included. Reactor sharing at the RCF brings outside groups into the facility for a tour, an explanation of reactor matters, and a reactor measurement. It has involved groups ranging from high school classes to advanced college groups and in size from a few to about 50 visitors. The RCF differs from other university reactors in that its fuel is like that of large power reactors, and its research and curriculum are dedicated to power reactor matters

  11. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

    2004-01-01

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  12. Pulsed reactors: A dissenting view

    International Nuclear Information System (INIS)

    Ganev, I.Kh.; Orlov, V.V.

    1995-01-01

    The preceding article, by G.A. Ivanov et al., contains interesting estimates of the expanded production of plutonium in thermonuclear explosions initiated by plutonium charges. It must be noted that more than 40 years of efforts, despite some technical successes, have not led to a fast-reactor technology suitable for large-scale power production. This explains the incessant search for a nuclear technology for the future and the renewed interest in accelerator, hybrid, and explosive approaches to plutonium production. The success of such efforts will depend largely on the formulation of goals and the choice of the principal criteria. It is appropriate to discuss these issues here because the adoption of the rate of plutonium production or the plutonium doubling time as the principal criterion sets the stage for the repetition of previous errors. However, as a preliminary, I would like to question some categorical assertions that were made by Ivanov et al. without the presentation of adequate supporting data (the assertions that open-quotes the creation of an power industry on the basis of ordinary breeder reactors is practically impossibleclose quotes and that open-quotes adequate power generation in the 21st centuryclose quotes is impossible). In fact, it is simple to calculate that, given a realistic doubling time for fast reactors of ∼10 years and the plutonium produced by thermal reactors (around 10 12 W), it would be possible, if so desired, to introduce power far exceeding 10 14 W in the 21st century

  13. Repetitively pulsed material testing facility

    International Nuclear Information System (INIS)

    Zucker, O.; Bostick, W.; Gullickson, R; Long, J.; Luce, J.; Sahlin, H.

    1975-01-01

    A continuously operated, 1 pps, dense-plasma-focus device capable of delivering a minimum of 10 15 neutrons per pulse for material testing purposes is described. Moderate scaling from existing results is sufficient to provide 2 x 10 13 n/cm 2 .s to a suitable target. The average power consumption, which has become a major issue as a result of the energy crisis, is analyzed with respect to other plasma devices and is shown to be highly favorable. A novel approach to the capacitor bank and switch design allowing repetitive operation is discussed. (U.S.)

  14. Repetitively pulsed material testing facility

    International Nuclear Information System (INIS)

    Zucker, O.; Bostick, W.; Gullickson, R.; Long, J.; Luce, J.; Sahlin, H.

    1975-01-01

    A continuously operated, 1 pps, dense-plasma-focus device capable of delivering a minimum of 10 15 neutrons per pulse for material testing purposes is described. Moderate scaling from existing results is sufficient to provide 2 x 10 13 n/cm 2 . s to a suitable target. The average power consumption, which has become a major issue as a result of the energy crisis, is analyzed with respect to other plasma devices and is shown to be highly favorable. A novel approach to the capacitor bank and switch design allowing repetitive operation is discussed

  15. SLAC pulsed X-ray facility

    Science.gov (United States)

    Ipe, N. E.; McCall, R. C.; Baker, E. D.

    1986-05-01

    The Stanford Linear Accelerator Center (SLAC) operates a high energy (up to 33 GeV) linear accelerator delivering pulses up to a few microseconds wide. The pulsed nature of the electron beam creates problems in the detection and measurement of radiation both from the accelerator beam and the klystrons that provide the RF power for the accelerator. Hence, a pulsed X-ray facility has been built at SLAC mainly for the purpose of testing the response of different radiation detection instruments to pulsed radiation fields. The X-ray tube consists of an electron gun with a control grid. This provides a stream of pulsed electrons that can be accelerated towards a confined target-window. The window is made up of aluminum 0.051 cm (20 mils) thick, plated on the vacuum side with a layer of gold 0.0006 cm (1/4 mil) thick. The frequency of electron pulses can be varied by an internal pulser from 60 to 360 pulses per second with pulse widths of 360 ns to 5 ms. The pulse amplitude can be varied over a wide range of currents. An external pulser can be used to obtain other frequencies or special pulse shapes. The voltage across the gun can be varied from 0 to 100 kV. The major part of the X-ray tube is enclosed in a large walk-in-cabinet made of 1.9 cm (3/4 in) plywood and lined with 0.32 cm (1/8 in) lead to make a very versatile facility.

  16. SLAC pulsed x-ray facility

    International Nuclear Information System (INIS)

    Ipe, N.E.; McCall, R.C.; Baker, E.D.

    1986-05-01

    The Stanford Linear Accelerator Center (SLAC) operates a high energy (up to 33 GeV) linear accelerator delivering pulses up to a few microseconds wide. The pulsed nature of the electron beam creates problems in the detection and measurement of radiation both from the accelerator beam and the klystrons that provide the rf power for the accelerator. Hence, a pulsed x-ray facility has been built at SLAC mainly for the purpose of testing the response of different radiation detection instruments to pulsed radiation fields. The x-ray tube consists of an electron gun with a control grid. This provides a stream of pulsed electrons that can be accelerated towards a confined target-window. The window is made up of aluminium 0.051 cm (20 mils) thick, plated on the vacuum side with a layer of gold 0.0006 cm (1/4 mil) thick. The frequency of electron pulses can be varied by an internal pulser from 60 to 360 pulses per second with pulse widths of 360 ns to 5 μs. The pulse amplitude can be varied over a wide range of currents. An external pulser can be used to obtain other frequencies or special pulse shapes. The voltage across the gun can be varied from 0 to 100 kV. The major part of the x-ray tube is enclosed in a large walk-in-cabinet made of 1.9 cm (3/4 in) plywood and lined with 0.32 cm (1/8 in) lead to make a very versatile facility. 3 refs., 5 figs

  17. Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Pickard, P.S.

    1976-01-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

  18. Design considerations for epithermal pulse reactors

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1978-01-01

    Simplified design criteria were developed for scoping analyses of epithermal pulse reactors for use in LMFBR safety testing. By using these criteria, materials and designs were investigated to determine performance limits of moderately sized reactor cores. Several designs are suggested for further study. These are a gas-cooled core fueled with a heterogeneous mixture of Fe-UO 2 cermet and BeO-UO 2 ceramic fuels, and a heavy-water-cooled core fueled with an Fe-UO 2 cermet

  19. STARFIRE remote maintenance and reactor facility concept

    International Nuclear Information System (INIS)

    Graumann, D.W.; Field, R.E.; Lutz, G.R.; Trachsel, C.A.

    1981-01-01

    A total remote maintenance facility has been designed for all equipment located within the reactor building and hot cell, although operational flexibility has been provided by design of the reactor shielding such that personnel access into the reactor building within 24 hours after reactor shutdown is possible. The reactor design permits removal and replacement of all components if necessary, however, the vacuum pumps, isolation valves and blanket require scheduled, routine maintenance. Reactor scheduled maintenance does not dominate annual plant downtime, therefore, several scheduled operations can be added without affecting reactor availability. The maintenance facilities consist of the reactor building, the hot cell, the reactor service area and the remote maintenance control room. The reactor building contains the reactor, selected support system modules, and required maintenance equipment. The reactor and the support systems are maintained with (1) equipment that is mounted on a monorail system; (2) overhead cranes; and (3) bridge-mounted electromechanical manipulators. The hot cell is located outside of the reactor building to localize contamination products and permit independent operation. An equipment air lock connects the reactor building to the hot cell

  20. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

  1. Alpha-contained laboratory scale pulse column facility for SRL

    International Nuclear Information System (INIS)

    Reif, D.J.; Cadieux, J.R.; Fauth, D.J.; Thompson, M.C.

    1980-01-01

    For studying solvent extraction processes, a laboratory-sized pulse column facility was constructed at the Savannah River Laboratory. This facility, in conjunction with existing miniature mixer-settler equipment and the centrifugal contactor facility currently under construction at SRL, provides capability for cross comparison of solvent extraction technology. This presentation describes the design and applications of the Pulse Column Facility at SRL

  2. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

    International Nuclear Information System (INIS)

    Dhami, P.S; Yadav, J.S; Agarwal, K.

    2017-01-01

    Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

  3. 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

  4. Pulse Star inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Blink, J.A.; Hogan, W.J.

    1985-01-01

    Pulse Star is a pool-type ICF reactor that emphasizes low cost and high safety levels. The reactor consists of a vacuum chamber (belljar) submerged in a compact liquid metal (Li 17 Pb 83 or lithium) pool which also contains the heat exchangers and liquid metal pumps. The shielding efficiency of the liquid metal pool is high enough to allow hands-on maintenance of (removed) pumps and heat exchangers. Liquid metal is allowed to spray through the 5.5 m radius belljar at a controlled rate, but is prohibited from the target region by a 4 m radius mesh first wall. The wetted first wall absorbs the fusion x-rays and debris while the spray region absorbs the fusion neutrons. The mesh allows vaporized liquid metal to blow through to the spray region where it can quickly cool and condense. Preliminary calculations show that a 2 m thick first wall could handle the mechanical (support, buckling, and x-ray-induced hoop) loads. Wetting and gas flow issues are in an initial investigation stage

  5. Claim prevention at reactor facilities

    International Nuclear Information System (INIS)

    Colby, B.P.

    1987-01-01

    Why does a radiation worker bring a claim alleging bodily injury from radiation exposure? Natural cancer, fear of radiation induced cancer, financial gain, emotional distress and mental anguish are some reasons for workers' claims. In this paper the author describes what power reactor health physicists are doing to reduce the likelihood of claims by establishing programs which provide sound protection of workers, prevent radiological events, improve workers' knowledge of radiological conditions and provide guidance for radiological incident response

  6. Effluent releases at the TRIGA reactor facility

    Energy Technology Data Exchange (ETDEWEB)

    Whittemore, W L [General Atomic Co., San Diego, CA (United States)

    1974-07-01

    The principal effluent from the operating TRIGA reactors in our facility is argon-41. As monitored by a recording gas and particulate stack monitor, the values shown in the table, the Mark III operating 24 hours per day for very long periods produced the largest amount of radioactive argon. The quantity of 23.7 Ci A-41 when diluted by the normal reactor room ventilation system corresponded to 1.45 x 10{sup -6} {mu}Ci/cc. As diluted in the roof stack stream and the reactor building wake, the concentration immediately outside the reactor building was 25% MPC for an unrestricted area. The continued dilution of this effluent resulted in a concentration of a few percent MPC at the site boundary (unrestricted area) 350 meters from the reactor. (author)

  7. 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

  8. Facilities of fuel transfer for nuclear reactors

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

  9. 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)

  10. Design and research of fuel element for pulsed reactor

    International Nuclear Information System (INIS)

    Tian Sheng

    1994-05-01

    The fuel element is the key component for pulsed reactor and its design is one of kernel techniques for pulsed reactor. Following the GA Company of US the NPIC (Nuclear Power Institute of China) has mastered this technique. Up to now, the first pulsed reactor in China (PRC-1) has been safely operated for about 3 years. The design and research of fuel element undertaken by NPIC is summarized. The verification and evaluation of this design has been carried out by using the results of measured parameters during operation and test of PRC-1 as well as comparing the design parameters published by others

  11. An electrical pulse hydride injector (EPHI) for reactor fueling and tritium handling applications

    International Nuclear Information System (INIS)

    Azizov, E.A.; Kareev, Yu.A.; Savotkin, A.N.; Frunze, V.V.; Penzhorn, R.D.; Glugla, M.

    1995-01-01

    An electrical pulse hydride injector (EPHI) has been developed for reactor fuelling as well as for handling of hydrogen isotopes in facilities operating with tritium. Salient features of the EPHI are the accuracy with which the fuelling rate can be controlled and the avoidance of a pressurized ballast. The generator is simple and allows for safe operation with tritium. (orig.)

  12. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

    2011-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  13. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  14. Pulsed irradiation of enriched UO{sub 2} in the Annular Core Pulse Reactor (ACPR)

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, T R; Lucoff, D M; Reil, K O; Croucher, D W [Sandia Laboratories (United States)

    1974-07-01

    A series of experiments have been conducted in the Annular Core Pulse Reactor (ACPR) to determine the energy deposition and behavior of enriched UO{sub 2} under pulse conditions. In the experiment single unirradiated pellets with enrichments up to 25 percent were pulse heated to melt temperatures. Temperature and fission product inventory measurements were made and compared with neutron transport calculations. (author)

  15. Cooling facility for reactor container

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro

    1996-05-31

    A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

  16. Cooling facility for reactor container

    International Nuclear Information System (INIS)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro.

    1996-01-01

    A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

  17. Reticulated Vitreous Carbon Electrodes for Gas Phase Pulsed Corona Reactors

    National Research Council Canada - National Science Library

    Locke, B

    1998-01-01

    A new design for gas phase pulsed corona reactors incorporating reticulated vitreous carbon electrodes is demonstrated to be effective for the removal of nitrogen oxides from synthetic air mixtures...

  18. Reticulated Vitreous Carbon Electrodes for Gas Phase Pulsed Corona Reactors

    National Research Council Canada - National Science Library

    LOCKE, B

    1999-01-01

    A new design for gas phase pulsed corona reactors incorporating reticulated vitreous carbon electrodes is demonstrated to be effective for the removal of nitrogen oxides from synthetic air mixtures...

  19. The investigation of enviromental radioactivity background around a pulsed reactor

    International Nuclear Information System (INIS)

    Xiao Tenghui; Zhao Zhongli

    1990-01-01

    The radioactivity background level of enviromental medium around a pulsed reactor for 5 km and external penetrating radioactivity dose level for 10 km are given. mediums measured include air, water, soil, organisms, fallout, etc

  20. The investigation of enviromental radioactivity background around a pulsed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tenghui, Xiao; Zhongli, Zhao [Southwest Inst. of Nuclear Reactor Engineering, Sichuan, SC (China)

    1990-06-01

    The radioactivity background level of enviromental medium around a pulsed reactor for 5 km and external penetrating radioactivity dose level for 10 km are given. mediums measured include air, water, soil, organisms, fallout, etc.

  1. Safety evaluation of the NSRR facility relevant to the modification for improved pulse operation and preirradiated fuel experiments

    International Nuclear Information System (INIS)

    Inabe, Teruo; Terakado, Yoshibumi; Tanzawa, Sadamitsu; Katagiri, Hiroshi; Kobayashi, Hideo

    1988-11-01

    The Nuclear Safety Research Reactor (NSRR) is a pulse reactor for the inpile experiments to study the fuel behavior under reactivity initiated accident conditions. The present operation modes of the NSRR consist of the steady state operation up to 300 kW and the natural pulse operation in which a sharp pulsed power is generated from substantially zero power level. In addition to these, two new modes of shaped pulse operation and combined pulse operation will be conducted in the near future as the improved pulse operations. A transient power up to 10 MW will be generated in the shaped pulse operation, and a combination of a transient power up to 10 MW and a sharp pulsed power will be generated in the combined pulse operation. Furthermore, preirradiated fuel rods will be employed in the future experiments whereas the present experiments are confined to the test specimens of unirradiated fuel rods. To provide for these programs, the fundamental design works relevant to the modification of the reactor facility including the reactor instrumentation and control systems and experimental provision were developed. The reactor safety evaluation is prerequisite for confirming the propriety of the fundamental design of the reactor facility from the safety point of view. The safety evaluation was therefore conducted postulating such events that would bring about abnormal conditions in the reactor facility. As a result of the safety evaluation, it has been confirmed as to the NSRR facility after modification that the anticipated transients, the postulated accidents, the major accident and the hypothetical accident do not result respectively in any serious safety problem and that the fundamental design principles and the reactor siting are adequate and acceptable. (author)

  2. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi

    2014-01-01

    Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

  3. Kinetic studies on a repetitively pulsed fast reactor

    International Nuclear Information System (INIS)

    Das, S.

    1982-01-01

    Neutronic analysis of an earlier proposed periodically pulsed fast reactor at Kalpakkam (KPFR) has been carried out numerically under equilibrium and transient conditions using the one-point model of reactor kinetics and the experimentally measured total worth of reactivity modulator, the parabolic coefficient of reactivity of the movable reflector and the mean prompt neutron lifetime. Results of steady-state calculations - treated on the basis of delayed neutron precursor and energy balances during a period of operation - have been compared with the analytical formulae of Larrimore for a parabolic reactivity input. Empirical relations for half-width of the fast neutron pulse, the peak pulse power and the power at first crossing of prompt criticality have been obtained and shown to be accurate enough for predicting steady-state power pulse characteristics of a periodically pulsed fast reactor. The concept of a subprompt-critical reactor has been used to calculate the fictitious delayed neutron fraction, β of the KPFR through a numerical experiment. Relative pulse height stability and pulse shape sensitivity to changes of maximum reactivity is discussed. With the aid of new safety concepts, the Power Amplification Factor (PAF) and the Pulse Growth Factor (Rsub(p)), the dynamics KPFR under accidental conditions has been studied for step and ramp reactivity perturbations. All the analysis has been done without taking account of reactivity feedback. (orig.)

  4. Pulsed lower-hybrid wave penetration in reactor plasmas

    International Nuclear Information System (INIS)

    Cohen, R.H.; Bonoli, P.T.; Porkolab, M.; Rognlien, T.D.

    1989-01-01

    Providing lower-hybrid power in short, intense (GW) pulses allows enhanced wave penetration in reactor-grade plasmas. We examine nonlinear absorption, ray propagation, and parametric instability of the intense pulses. We find that simultaneously achieving good penetration while avoiding parametric instabilities is possible, but imposes restrictions on the peak power density, pulse duration, and/or r.f. spot shape. In particular, power launched in narrow strips, elongated along the field direction, is desired

  5. Recent operational history of the new Sandia Pulsed Reactor III (SPR III)

    International Nuclear Information System (INIS)

    Schmidt, T.R.; Estes, B.F.; Reuscher, J.A.

    1977-01-01

    The Sandia Pulsed Reactor III (SPR III) is a fast-pulse research reactor which was designed and built at Sandia Laboratories and achieved criticality in August 1975. The reactor is now characterized and is in an operational configuration. The core consists of 18 fuel plates (258 kg fuel mass) of fully enriched uranium alloyed with 10 wt.% molybdenum. It is arranged in an annular configuration with an inside diameter of 17.78 cm, an outside diameter of 29.72 cm, and a height of 35.9 cm. The reactor core uses reflectors of copper and aluminum for control and an external bolting arrangement to secure the fuel plates. SPR III and SPR II are operated on an interchangeable basis using the same facility and control system. As of June 1977, SPR III has had over 240 operations with core temperatures up to 541 0 C

  6. Emergency facility control device for nuclear reactor

    International Nuclear Information System (INIS)

    Ikehara, Morihiko.

    1981-01-01

    Purpose: To increase the reliability of a nuclear reactor by allowing an emergency facility to be manually started and stopped to make its operation more convenient and eliminate the possibility of erroneous operation in an emergency. Constitution: There are provided a first water level detector for detecting a level lower than the first low water level in a reactor container and a second water level detector for detecting a level lower than the second low water level lower than the first low water level, and an emergency facility can be started and stopped manually only when the level is higher than the second low water level, but the facility will be started regardless of the state of the manual operation when the level is lower than the second low water level. Thus, the emergency facility can be started by manual operation, but will be automatically started so as to secure the necessary minimum operation if the level becomes lower than the second low water level and the stopping operation thereafter is forgotten. (Kamimura, M.)

  7. Pilot-scale reactor activation facility at SRL

    International Nuclear Information System (INIS)

    Bowman, W.W.

    1976-01-01

    The Hydrogeocemical and Stream Sediment Reconnaissance portion of the National Uranium Resource Evaluation program requires an analytical technique for uranium and other elements. Based on an automated absolute activation analysis technique using 252 Cf, a pilt-scale facility installed in a production reactor has provided analyses for 2800 samples. Key features include: an automated sample transport system, a delayed neutron detector, two GeLi detectors, a loader, and an unloader, with all components controlled by a microprocessor; a dedicated PDP-9 computer and pulse height analyzer; and correlation and reduction of acquired data by a series of programs using an IBM 360/195 computer. The facility was calibrated with elemental and isotopic standards. Results of analyses of standard reference materials and operational detection limits for typical sediment samples are presented. Plans to increase sample throughput are discussed briefly

  8. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  9. Study of startup conditions of a pulsed annular reactor

    International Nuclear Information System (INIS)

    Silva, Mario Augusto Bezerra da

    2003-10-01

    A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

  10. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

    2013-06-01

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  11. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

    In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

  12. 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

  13. Risk management activities at the DOE Class A reactor facilities

    International Nuclear Information System (INIS)

    Sharp, D.A.; Hill, D.J.; Linn, M.A.; Atkinson, S.A.; Hu, J.P.

    1993-01-01

    The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented

  14. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Quapp, W.J.; Watts, K.D.

    1985-01-01

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  15. Forge: a short pulse x-ray diagnostic development facility

    International Nuclear Information System (INIS)

    Stradling, G.L.; Hurry, T.R.; Denbow, E.R.; Selph, M.M.; Ameduri, F.P.

    1985-01-01

    A new short pulse x-ray calibration facility has been brought on line at Los Alamos. This facility is being used for the development, testing and calibration of fast x-ray diagnostic systems. The x-ray source consists of a moderate size, sub-nanosecond laser focused at high intensity on an appropriate target material to generate short pulses of x-ray emission from the resulting plasma. Dynamic performance parameters of fast x-ray diagnostic instruments, such as x-ray streak cameras, can be conveniently measured using this facility

  16. Assessment and mitigation of electromagnetic pulse (EMP) impacts at short-pulse laser facilities

    International Nuclear Information System (INIS)

    Brown, C G Jr; Bond, E; Clancy, T; Dangi, S; Eder, D C; Ferguson, W; Kimbrough, J; Throop, A

    2010-01-01

    The National Ignition Facility (NIF) will be impacted by electromagnetic pulse (EMP) during normal long-pulse operation, but the largest impacts are expected during short-pulse operation utilizing the Advanced Radiographic Capability (ARC). Without mitigation these impacts could range from data corruption to hardware damage. We describe our EMP measurement systems on Titan and NIF and present some preliminary results and thoughts on mitigation.

  17. Effects of pulse-to-pulse residual species on discharges in repetitively pulsed discharges through packed bed reactors

    Science.gov (United States)

    Kruszelnicki, Juliusz; Engeling, Kenneth W.; Foster, John E.; Kushner, Mark J.

    2016-09-01

    Atmospheric pressure dielectric barrier discharges (DBDs) sustained in packed bed reactors (PBRs) are being investigated for conversion of toxic and waste gases, and CO2 removal. These discharges are repetitively pulsed having varying flow rates and internal geometries, which results in species from the prior pulse still being in the discharge zone at the time the following discharge pulse occurs. A non-negligible residual plasma density remains, which effectively acts as preionization. This residual charge changes the discharge properties of subsequent pulses, and may impact important PBR properties such as chemical selectivity. Similarly, the residual neutral reactive species produced during earlier pulses will impact the reaction rates on subsequent pulses. We report on results of a computational investigation of a 2D PBR using the plasma hydrodynamics simulator nonPDPSIM. Results will be discussed for air flowing though an array of dielectric rods at atmospheric pressure. The effects of inter-pulse residual species on PBR discharges will be quantified. Means of controlling the presence of residual species in the reactor through gas flow rate, pulse repetition, pulse width and geometry will be described. Comparisons will be made to experiments. Work supported by US DOE Office of Fusion Energy Science and the National Science Foundation.

  18. A remote maintenance robot system for a pulsed nuclear reactor

    International Nuclear Information System (INIS)

    Thunborg, S.

    1987-01-01

    This paper presents a remote maintenance robot system for use in a hazardous environment. The system consists of turntable, robot and hoist subsystems which operate under the control of a supervisory computer to perform coordinated programmed maintenance operations on a pulsed nuclear reactor. The system is operational

  19. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    Salazar, M.; Elder, J.

    1992-08-01

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program

  20. Magnet design approach for pulsed tokamak reactors

    International Nuclear Information System (INIS)

    Kim, S.H.; Evans, K. Jr.; Ehst, D.A.

    1983-12-01

    A choice of various operating modes of a tokamak reactor will have considerable impact on the fatigue lives and cost of ohmic heating (OH), equilibrium field (EF), and toroidal field (TF) coils. OH AND EF coil requirements and their costs, as well as the effects of the fringing fields of the EF coils on the TF coils, have been studied under cyclic operation in the range of N = 10 2 to 10 6 cycles, spanning the range from a noninductively driven reactor (STARFIRE) to a conventional ohmically driven reactor. For a reference design of TF coils the design of the central OH solenoid has been studied as a function of its maximum field, B/sup OH/. Increasing requirements for structural support lead to only negligible increases in volt-seconds for B/sup OH/ greater than or equal to 10.0 T. Fatigue failure of the OH coil is not a concern for N less than or equal to 10 5 ; for N approx. 10 6 fatigue limits the strain to small values, resulting in small increases in structural requirements and modest decreases in volt-seconds. Should noninductive current drive be achievable we note that this not only eliminates the OH coil, but it also permits EF coil placement in the inboard region, which facilitates the creation of highly shaped plasma cross sections (large triangularity, or bean-shaped equilibria). We have computed the stored energy, coil configuration and fringing fields for a number of EF coil design options

  1. Development of the pulsed muon facility at ISIS

    International Nuclear Information System (INIS)

    Eaton, G.H.; Scott, C.A.; Williams, W.G.

    1994-01-01

    The ISIS pulsed surface muon facility at RAL is presently undergoing a major expansion to provide three experimental ports with simultaneous single muon pulses at 50 Hz. This upgrade, funded by the European Community (EC), is described together with recent development results which are relevant to its future scientific programme. These new beam lines are expected to be available for experiments in June 1993. (orig.)

  2. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  3. Present status of decommissioning in the Musashi Reactor Facility (4)

    International Nuclear Information System (INIS)

    Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

    2012-01-01

    The decommissioning of the Musashi reactor was decided in 2003. Permanent shutdown of the reactor and stopping the operational functions were conducted in 2004. Transportation of the spent fuels was finished in 2006. After 2007, the system and equipment stopping the functions were stored as installed in the reactor facility as radioactive wastes. After separating nonradioactive wastes such as concretes from radioactive wastes with a contamination test, stopping the functions of liquid waste management facility was performed with newly installed drainage facility for radioisotope use in 2010. Solid waste management facility was also dismantled and removed in the same way as liquid waste management facility in 2011. Radioactive wastes packed in containers were moved and stored in the reactor facility. (T. Tanaka)

  4. Investigation of the Pulsed Annular Gas Jet for Chemical Reactor Cleaning

    Directory of Open Access Journals (Sweden)

    Zvegintsev Valery Ivanovich

    2012-01-01

    Full Text Available The most economical technology for production of titanium dioxide pigment is plasma-chemical syntheses with the heating of the oxygen. The highlight of the given reaction is formation of a solid phase as a result of interactions between two gases, thus brings the formation of particle deposits on the reactor walls, and to disturbing the normal operation of the technological process. For the solving of the task of reactor internal walls cleaning the pulsed gaseous system was suggested and investigated, which throws circular oxygen jet along surfaces through regular intervals. Study of aerodynamic efficiency of the impulse system was carried by numerical modeling and experimentally with the help of a specially created experimental facility. The distribution of the pulsed flow velocity at the exit of cylindrical reactor was measured. The experimental results have shown that used impulse device creates a pulsed jet with high value of the specified flow rate. It allows to get high velocities that are sufficient for the particle deposits destruction and their removal away. Designed pulsed peelings system has shown high efficiency and reliability in functioning that allows us to recommend it for wide spreading in chemical industry.

  5. Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR.

    Science.gov (United States)

    Mitchell, G E; Furman, W I; Lychagin, E V; Muzichka, A Yu; Nekhaev, G V; Strelkov, A V; Sharapov, E I; Shvetsov, V N; Chernuhin, Yu I; Levakov, B G; Litvin, V I; Lyzhin, A E; Magda, E P; Crawford, B E; Stephenson, S L; Howell, C R; Tornow, W

    2005-01-01

    Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 10(18)/cm(2)s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286.

  6. Characterization of graphite-matrix pulsed reactor fuels

    International Nuclear Information System (INIS)

    Karnes, C.H.; Marion, R.H.

    1976-01-01

    The performance of the Annular Core Pulsed Reactor (ACPR) is being upgraded in order to accommodate higher fluence experiments for fast reactor fuel element transient and safety studies. The increased fluence requires a two-zone core with the inner zone containing fuel having a high enthalpy and the capability of withstanding very high temperatures during both pulsed and steady state operation. Because the fuel is subjected to a temperature risetime of 2 to 5 ms and to a large temperature difference across the diameter, fracture due to thermal stresses is the primary failure mode. One of the fuels considered for the high enthalpy inner region is a graphite-matrix fuel containing a dispersion of uranium--zirconium carbide solid solution particles. A program was initiated to optimize the development of this class of fuel. This summary presents results on formulations of fuel which have been fabricated by the Materials Technology Group of the Los Alamos Scientific Laboratory

  7. Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

    International Nuclear Information System (INIS)

    1995-01-01

    With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor's Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced

  8. Pulse generation and preamplification for long pulse beamlines of Orion laser facility.

    Science.gov (United States)

    Hillier, David I; Winter, David N; Hopps, Nicholas W

    2010-06-01

    We describe the pulse generation, shaping, and preamplification system for the nanosecond beamlines of the Orion laser facility. The system generates shaped laser pulses of up to approximately 1 J of 100 ps-5 ns duration with a programmable temporal profile. The laser has a 30th-power supergaussian spatial profile and is diffraction limited. The system is capable of imposing 2D smoothing by spectral dispersion upon the beam, which will produce a nonuniformity of 10% rms at the target.

  9. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-07-01

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

  11. Magnox Electric Littlebrook reactor inspection and repair rehearsal facility

    International Nuclear Information System (INIS)

    Barnes, S.A.; Clayton, R.; Gaydon, B.G.; Ramsey, B.H.

    1996-01-01

    Magnox reactors, although designed to be maintenance free during their operational life, have nevertheless highlighted the need for test rig facilities to train operators in the methods and techniques of reactor inspection and repair. The history of the facility for reactor engineering development (FRED) is described and its present role as a repair rehearsal facility noted. Advances in computer graphics may, in future, mean that such operator training will be virtual reality rather than analog reality based; however the need for such rigs to commission techniques and equipment and to establish performance and reliability is likely to continue. (UK)

  12. Circuit designs for measuring reactor period, peak power, and pulse fluence on TRIGA and other pulse reactor

    International Nuclear Information System (INIS)

    Meyer, R.D.; Thome, F.V.; Williams, R.L.

    1976-01-01

    Inexpensive circuits for use in evaluating reactor pulse prompt period, peak power, and pulse fluence (NVT) are presented. In addition to low cost, these circuits are easily assembled and calibrated and operate with a high degree of accuracy. The positive period measuring system has been used in evaluating reactivity additions as small as 5 cents (with an accuracy of ±0.1 cents) and as large as $4.50 (accuracy ±2 cents). Reactor peak power is measured digitally with a system accuracy of ±0.04% of a 10 Volt input (±4 mV). The NVT circuit measures over a 2-1/2 decade range, has 3 place resolution and an accuracy of better than 1%. (author)

  13. General considerations for neutron capture therapy at a reactor facility

    International Nuclear Information System (INIS)

    Binney, S.E.

    2001-01-01

    In addition to neutron beam intensity and quality, there are also a number of other significant criteria related to a nuclear reactor that contribute to a successful neutron capture therapy (NCT) facility. These criteria are classified into four main categories: Nuclear design factors, facility management and operations factors, facility resources, and non-technical factors. Important factors to consider are given for each of these categories. In addition to an adequate neutron beam intensity and quality, key requirements for a successful neutron capture therapy facility include necessary finances to construct or convert a facility for NCT, a capable medical staff to perform the NCT, and the administrative support for the facility. The absence of any one of these four factors seriously jeopardizes the overall probability of success of the facility. Thus nuclear reactor facility management considering becoming involved in neutron capture therapy, should it be proven clinically successful, should take all these factors into consideration. (author)

  14. Material accountancy and control practice at a research reactor facility

    International Nuclear Information System (INIS)

    Bouchard, J.; Maurel, J.J.; Tromeur, Y.

    1982-01-01

    This session surveys the regulations, organization, and accountancy practice that compose the French State System of Accountancy and Control. Practical examples are discussed showing how inventories are verified at a critical assembly facility and at a materials testing reactor

  15. Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

    International Nuclear Information System (INIS)

    1995-01-01

    The RPI Reactor Critical Facility (RCF) operated successfully over the period fall 1994 - fall 1995. During this period, the RCF was used for Critical Reactor Laboratory spring 1995 (12 students); Reactor Operations Training fall 1994 (3 students); Reactor Operations Training spring 1995 (3 students); and Reactor Operations Training fall 1995 (3 students). Thirty-two Instrumentation and Measurement students used the RCF for one class for hands-on experiments with nuclear instruments. In addition, a total of nine credits of PhD thesis work were carried out at the RCF. This document constitutes the 1995 Report of the Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) to the USNRC, to the USDOE, and to RPI management

  16. Brookhaven Reactor Experiment Control Facility, a distributed function computer network

    International Nuclear Information System (INIS)

    Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

    1975-11-01

    A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented

  17. Progress report concerning safety research for nuclear reactor facilities

    International Nuclear Information System (INIS)

    1978-01-01

    Examination and evaluation of safety research results for nuclear reactor facilities have been performed, as more than a year has elapsed since the plan had been initiated in April, 1976, by the special sub-committee for the safety of nuclear reactor facilities. The research is carried out by being divided roughly into 7 items, and seems to be steadily proceeding, though it does not yet reach the target. The above 7 items include researches for (1) criticality accident, (2) loss of coolant accident, (3) safety for light water reactor fuel, (4) construction safety for reactor facilities, (5) reduction of release of radioactive material, (6) safety evaluation based on the probability theory for reactor facilities, and (7) aseismatic measures for reactor facilities. With discussions on the progress and the results of the research this time, research on the behaviour on fuel in abnormal transients including in-core and out-core experiments has been added to the third item, deleting the power-cooling mismatch experiment in Nuclear Safety Research Reactor of JAERI. Also it has been decided to add two research to the seventh item, namely measured data collection, classification and analysis, and probability assessment of failures due to an earthquake. For these 7 items, the report describes the concrete contents of research to be performed in fiscal years of 1977 and 1978, by discussing on most rational and suitable contents conceivable at present. (Wakatsuki, Y.)

  18. 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.)

  19. Efficient modeling for pulsed activation in inertial fusion energy reactors

    International Nuclear Information System (INIS)

    Sanz, J.; Yuste, P.; Reyes, S.; Latkowski, J.F.

    2000-01-01

    First structural wall material (FSW) materials in inertial fusion energy (IFE) power reactors will be irradiated under typical repetition rates of 1-10 Hz, for an operation time as long as the total reactor lifetime. The main objective of the present work is to determine whether a continuous-pulsed (CP) approach can be an efficient method in modeling the pulsed activation process for operating conditions of FSW materials. The accuracy and practicability of this method was investigated both analytically and (for reaction/decay chains of two and three nuclides) by computational simulation. It was found that CP modeling is an accurate and practical method for calculating the neutron-activation of FSW materials. Its use is recommended instead of the equivalent steady-state method or the exact pulsed modeling. Moreover, the applicability of this method to components of an IFE power plant subject to repetition rates lower than those of the FSW is still being studied. The analytical investigation was performed for 0.05 Hz, which could be typical for the coolant. Conclusions seem to be similar to those obtained for the FSW. However, further future work is needed for a final answer

  20. Long-pulse magnetic field facility at Zaragoza

    International Nuclear Information System (INIS)

    Algarabel, P A; Moral, A del; Martin, C; Serrate, D; Tokarz, W

    2006-01-01

    The long-pulse magnetic field facility of the Laboratorio de Magnetismo - Instituto de Ciencia de Materiales de Aragon (Universidad de Zaragoza-CSIC) produces magnetic fields up to 31, with a pulse duration of 2.2s. Experimental set-ups for measurements of magnetization, magnetostriction and magnetoresistance are available. The temperature can be controlled between 1.4 and 335 K, being the inner bore of the He cryostat of 22.5 mm. Magnetization is measured using the mutual induction technique, the magnetostriction is determined with the strain-gage and the capacitive cantilever methods, and the magnetoresistance is measured by means of the aclock-in technique in the 4-probes geometry. An overview of the facility will be presented and the presently available experimental techniques will be discussed

  1. Long-pulse magnetic field facility at Zaragoza

    Science.gov (United States)

    Algarabel, P. A.; del Moral, A.; Martín, C.; Serrate, D.; Tokarz, W.

    2006-11-01

    The long-pulse magnetic field facility of the Laboratorio de Magnetismo - Instituto de Ciencia de Materiales de Aragón (Universidad de Zaragoza-CSIC) produces magnetic fields up to 31, with a pulse duration of 2.2s. Experimental set-ups for measurements of magnetization, magnetostriction and magnetoresistance are available. The temperature can be controlled between 1.4 and 335 K, being the inner bore of the He cryostat of 22.5 mm. Magnetization is measured using the mutual induction technique, the magnetostriction is determined with the strain-gage and the capacitive cantilever methods, and the magnetoresistance is measured by means of the aclock-in technique in the 4-probes geometry. An overview of the facility will be presented and the presently available experimental techniques will be discussed.

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

    International Nuclear Information System (INIS)

    Liu Jianjun; Zhou Zhigao; Zhang Ben'ai

    2007-01-01

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

  3. Design Guide for Category I reactors critical facilities

    International Nuclear Information System (INIS)

    Brynda, W.J.; Powell, R.W.

    1978-08-01

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned critical facilities be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission

  4. AIREK-PUL, Periodic Kinetics Problems of Pulsed Reactors

    International Nuclear Information System (INIS)

    Inzaghi, A.; Misenta, R.

    1984-01-01

    1 - Nature of physical problem solved: Solves periodic problems about the kinetics of pulsed reactors or problems where the reactivity has rapid variations. The program uses two constant steps for the integration of the system of differential equations, the first step during the first half-period and the second step during the second half-period. Available for either single or double precision. 2 - Method of solution: The differential equations are integrated using the fourth-order Runge-Kutta method as modified by E.R. Cohen (Geneva Conference, 1958). 3 - Restrictions on the complexity of the problem: The maximum number of differential equations that can be solved simultaneously is 50

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

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

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

  6. Facility with a nuclear district heating reactor

    International Nuclear Information System (INIS)

    Straub, H.

    1988-01-01

    The district heating reactor has a pressure vessel which contains the reactor core and at least one coolant conducting primary heat carrier surrounded by a heat sink. The pressure vessel has two walls with a space between them. This space is connected with a container which contains air as heat isolating medium and water as heat conducting medium. During the normal reactor operation the space is filled by air from the container with the aid of a blower, whereas in the case of a break-down of the cooling system it is filled by water which flows out of the container by gravity after the blower has been switched off. The after-heat, generated in the reactor core during cooling break-down, is removed into the heat sink surrounding the pressure vessel in a safe and simple way. 6 figs

  7. MHD stability regimes for steady state and pulsed reactors

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kessel, C.E.; Pomphrey, N.

    1994-02-01

    A tokamak reactor will operate at the maximum value of β≡2μ 0 /B 2 that is compatible with MHD stability. This value depends upon the plasma current and pressure profiles, the plasma shape and aspect ratio, and the location of nearby conducting structures. In addition, a steady state reactor will minimize its external current drive requirements and thus achieve its maximum economic benefit with a bootstrap fraction near one, I bs /I p ∼ 1, which constrains the product of the inverse aspect ratio and the plasma poloidal beta to be near unity, ε β p ∼ 1. An inductively driven pulsed reactor has different constraints set by the steady-state Ohm's law which relates the plasma temperature and density profiles to the parallel current density. We present the results obtained during the ARIES I, II/IV, and III and the PULSAR reactor studies where these quantities were optimized subject to different design philosophies. The ARIES-II/IV and ARIES-III designs are both in the second stability regime, but differ in requirements on the form of the profiles at the plasma edge, and in the location of the conducting wall. The relation between these, as well as new attractive MHD regimes not utilized in the ARIES or PULSAR studies is also discussed

  8. Magnetohydrodynamic stability regimes for steady state and pulsed reactors

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kessel, C.E.; Pomphrey, N.

    1994-01-01

    A tokamak reactor will operate at the maximum value of β≡2μ 0 left angle p right angle /B 2 that is compatible with magnetohydrodynamic (MHD) stability. This value depends on the plasma current and pressure profiles, the plasma shape and aspect ratio, and the location of nearby conducting structures. In addition, a steady state reactor will minimize its external current drive requirements and thus achieve its maximum economic benefit with a bootstrap fraction near unity, I BS /I P ∼1, which constrains the product of the inverse aspect ratio and the plasma poloidal β to be near unity, arepsilonβ P ∼1. An inductively driven pulsed reactor has different constraints set by the steady-state Ohm's law which relates the plasma temperature and density profiles to the parallel current density. We present the results obtained during ARIES I, II/IV, and III and PULSAR reactor studies where these quantities were optimized subject to different design philosophies. The ARIES-II/IV and ARIES-III designs are both in the second stability regime, but differ in requirements in the form of the profiles at the plasma edge, and in the location of the conducting wall. The relation between these, as well as new attractive MHD regimes not utilized in the ARIES or PULSAR studies, is also discussed. ((orig.))

  9. Coolant cleaning facility for nuclear reactor

    International Nuclear Information System (INIS)

    Kuboniwa, Takao; Konno, Yasuhiro; Kumaya, Shin; Osumi, Katsumi.

    1982-01-01

    Purpose: To remove cation of radioactive cobalt 60 produced in a reactor water during the ordinary operation of the reactor and chlorine when sea water is leaked in a condenser as well as to suppress an increase in iron clad containing radioactive cobalt 60 in the reactor water when the reactor is stopped. Constitution: A large flow rate high temperature cleaning system having an electromagnetic filter capable of removing radioactive substance in a reactor water, a low temperature cleaning system having a desalting unit using ion exchanger resin, a turbidity meter for measuring the turbidity of the reactor water and a conductivity meter for measuring the conductivity are provided. Further, flow rate control means are provided in the high and low temperature cleaning systems. The flow rate control means of the high temperature cleaning system is controlled by a measured signal of the turbidity meter, and the flow rete control means of the low temperature cleaning system is controlled by the measured signal of the conductivity meter. (Aizawa, K.)

  10. Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

    Energy Technology Data Exchange (ETDEWEB)

    Gutmann, H

    1972-06-05

    The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

  11. Coupling of AST-500 heating reactors with desalination facilities

    International Nuclear Information System (INIS)

    Kourachenkov, A.V.

    1998-01-01

    The general issues regarding NHR and desalination facility joint operation for potable water production are briefly considered. AST-500 reactor plant and DOU GTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. Similarity of NHR operation for a heating grid and a desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author)

  12. Coupling of AST-500 heating reactors with desalination facilities

    International Nuclear Information System (INIS)

    Gureyeva, L.V.; Egorov, V.V.; Podberezniy, V.L.

    1997-01-01

    The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab

  13. Coupling of AST-500 heating reactors with desalination facilities

    Energy Technology Data Exchange (ETDEWEB)

    Gureyeva, L V; Egorov, V V [OKBM, Nizhny Novgorod (Russian Federation); Podberezniy, V L [Scientific Research Inst. of Machine Building, Ekaterinburg (Russian Federation)

    1997-09-01

    The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab.

  14. Design requirements for new nuclear reactor facilities in Canada

    International Nuclear Information System (INIS)

    Shim, S.; Ohn, M.; Harwood, C.

    2012-01-01

    The Canadian Nuclear Safety Commission (CNSC) has been establishing the regulatory framework for the efficient and effective licensing of new nuclear reactor facilities. This regulatory framework includes the documentation of the requirements for the design and safety analysis of new nuclear reactor facilities, regardless of size. For this purpose, the CNSC has published the design and safety analysis requirements in the following two sets of regulatory documents: 1. RD-337, Design of New Nuclear Power Plants and RD-310, Safety Analysis for Nuclear Power Plants; and 2. RD-367, Design of Small Reactor Facilities and RD-308, Deterministic Safety Analysis for Small Reactor Facilities. These regulatory documents have been modernized to document past practices and experience and to be consistent with national and international standards. These regulatory documents provide the requirements for the design and safety analysis at a high level presented in a hierarchical structure. These documents were developed in a technology neutral approach so that they can be applicable for a wide variety of water cooled reactor facilities. This paper highlights two particular aspects of these regulatory documents: The use of a graded approach to make the documents applicable for a wide variety of nuclear reactor facilities including nuclear power plants (NPPs) and small reactor facilities; and, Design requirements that are new and different from past Canadian practices. Finally, this paper presents some of the proposed changes in RD-337 to implement specific details of the recommendations of the CNSC Fukushima Task Force Report. Major changes were not needed as the 2008 version of RD-337 already contained requirements to address most of the lessons learned from the Fukushima event of March 2011. (author)

  15. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  16. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  17. Emergency water supply facility for nuclear reactor

    International Nuclear Information System (INIS)

    Karasawa, Toru

    1998-01-01

    Water is stored previously in an equipment storage pit disposed on an operator floor of a reactor building instead of a condensate storage vessel. Upon occurrence of an emergency, water is supplied from the equipment storage pit by way of a sucking pipeline to a pump of a high pressure reactor core water injection circuit and a pump of a reactor-isolation cooling circuit to supply water to a reactor. The equipment storage pit is arranged in a building so that the depth thereof is determined to keep the required amount of water by storing water at a level lower than the lower end of a pool gate during normal operation. Water is also supplied from the equipment storage pit by way of a supply pipeline to a spent fuel storage pool on the operation floor of the reactor building. Namely, water is supplied to the spent fuel storage pool by a pump of a fuel pool cooling and cleaning circuit. This can eliminate a suppression pool cleaning circuit. (I.N.)

  18. Standard irradiation facilities for use in TRIGA reactors

    International Nuclear Information System (INIS)

    Kolbasov, B.N.; Luse, R.A.

    1972-01-01

    The standard neutron irradiation facility (SNIP) was developed under IAEA and FAO co-ordinated research program for the standardization of neutron irradiation facilities for radiobiological research, resulting in the possibility to use fast neutrons from pool-type reactors for radiobiological studies. The studies include irradiation of seeds for crop improvement, of Drosophila for genetic studies, and of microorganisms for developing industrially useful mutants, as well as fundamental studies in radiation biology. The facilities, located in the six pool-type reactors (in Austria, Bulgaria, India, Philippines, Thailand and Taiwan), have been calibrated and utilized to compare the response to fast neutrons of barley seeds (variety Himalaya CI 000620) which were selected as a standard biological monitor by which to estimate neutron fluxes in different reactors. These comparative irradiation studies showed excellent agreement and reproducibility

  19. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  20. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  1. Neutron beam facilities at the replacement research reactor

    International Nuclear Information System (INIS)

    Kennedy, S.

    1999-01-01

    Full text: On September 3rd 1997 the Australian Federal Government announced their decision to replace the HIFAR research reactor by 2005. The proposed reactor will be a multipurpose reactor with improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The neutron beam facilities are intended to cater for Australian scientific needs well into the 21st century. In the first stage of planning the neutron Beam Facilities at the replacement reactor, a Consultative Group was formed (BFCG) to determine the scientific capabilities of the new facility. Members of the group were drawn from academia, industry and government research laboratories. The BFCG submitted their report in April 1998, outlining the scientific priorities to be addressed. Cold and hot neutron sources are to be included, and cold and thermal neutron guides will be used to position most of the instruments in a neutron guide hall outside the reactor confinement building. In 2005 it is planned to have eight instruments installed with a further three to be developed by 2010, and seven spare instrument positions for development of new instruments over the life of the reactor. A beam facilities technical group (BFTG) was then formed to prepare the engineering specifications for the tendering process. The group consisted of some members of the BFCG, several scientists and engineers from ANSTO, and scientists from leading neutron scattering centres in Europe, USA and Japan. The BFTG looked in detail at the key components of the facility such as the thermal, cold and hot neutron sources, neutron collimators, neutron beam guides and overall requirements for the neutron guide hall. The report of the BFTG, completed in August 1998, was incorporated into the draft specifications for the reactor project, which were distributed to potential reactor vendors. An assessment of the first stage of reactor vendor submissions was completed in

  2. Proposal for a seismic facility for reactor safety research

    International Nuclear Information System (INIS)

    Anderson, C.A.; Dove, R.C.; Rhorer, R.L.

    1976-07-01

    Certain problem areas in the seismic analysis and design of nuclear reactors are enumerated and the way in which an experimental program might contribute to each area is examined. The use of seismic simulation testing receives particular attention, especially with regard to the verification of structural response analysis. The importance of scale modeling used in conjunction with seismic simulation is also stressed. The capabilities of existing seismic simulators are summarized, and a proposed facility is described which would considerably extend the ability to conduct, with confidence, confirmatory experiments on the behavior of reactor components when subjected to seismic excitation. Particular applications to gas-cooled and other reactor types are described

  3. Pulsed Nd-YAG laser welding of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Suresh Varma, P.V.; Gupta, Amit; Amit, K.; Bhatt, R.B.; Afzal, Mohd.; Panakkal, J.P.; Kamath, H.S.

    2009-02-01

    End plug welding of Prototype Fast Breeder Reactor (PFBR) fuel elements involves welding of fully Austenitic Stainless Steel (ASS) of grade D9 clad tube with 316M end plug. Pulsed Gas Tungsten Arc Welding (GTAW) is being used for the production of PFBR fuel elements at Advanced Fuel Fabrication Facility (AFFF). GTAW is an established process for end plug welding and hence adopted by many countries. GTAW has got certain limitations like heat input, arc gap sensitivity and certain sporadic defects like tungsten inclusion. Experiments have been carried out at AFFF to use Laser Beam Welding (LBW) technique as LBW offers a number of advantages over the former process. This report mainly deals with the optimization of laser parameters for welding of PFBR fuel elements. To facilitate pulsed Nd-YAG laser spot welding, parameters like peak power, pulse duration, pulse energy, frequency and defocusing of laser beam on to the work piece have been optimized. On the basis of penetration requirement laser welding parameters have been optimized. (author)

  4. A pulsed fast reactor; Un reacteur pulse a neutrons rapides; Impul'snyj reaktor na bystrykh nejtronakh; Reactor rapido pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, G. E.; Blokhintsev, D. I.; Blyumkina, Yu. A.; Bondarenko, I. I.; Deryagin, B. N.; Zajmovskij, A. S.; Zinov' ev, V. P.; Kazachkovskij, O. D.; Krasnoyarov, N. V.; Lejpunskij, A. I.; Malykh, V. A.; Nazarov, P. M.; Nikolaev, S. K.; Stavisskij, Yu. Ya.; Ukraintsev, F. I.; Frank, I. M.; Shapiro, F. Ji.; Yazvitskij, Yu. S. [Akademiya Nauk, Moscow, SSSR (Russian Federation)

    1962-03-15

    A pulsed fast reactor (IBR) has been operating at rated capacity since December 1960 in the Joint Institute for Nuclear Research. This reactor is used as a pulsed neutron source for physical experiments carried out by the time-of-flight method. It is used for total cross-section and intermediate neutron capture cross- section measurements, for studying the interaction between slow neutrons and solids and liquids, and for measuring neutron spectra produced in various media. The paper describes the basic structural features of the reactor and the results of the experiments for which it has been used. The reactor's operating system is based on recurrent pulses. Power pulses are produced when the mobile part of the reactor core moves swiftly through the stationary part of the core. The mobile part of the core is fastened to a rotating disc and travels at a speed of 230 m/s. The frequency of power pulses can be altered by means of an auxiliary mobile zone which has a range of 2.3-88 pulses per second. The mean power of the reactor is 1 kW, and the half-width of the power pulse in 36 {mu}s. The reactor is provided with a control and safety system which ensures automatic maintenance of mean power and swift shutdown in the event of any operational irregularity. It is fitted with a system of evacuated-neutron-flight tubes used in time-of-flight experiments. The main tube is 1000 m in length. In the start-up process and during physical experiments carried out on the reactor, the influence on reactivity of displacing the controls and the mobile parts of the core was studied ; the length of the pulse was measured under various operating conditions, and power pulse amplitude fluctuations were studied. Further measurements were made to establish the lifetime of prompt neutrons, the effective fraction of delayed neutrons, and coefficients of reactivity. (author) [French] L'Institut unifie de recherches nucleaires dispose d'un reacteur puise a neutrons rapides (IBR), qui

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  6. Gamma ray thermometrical facility for nuclear reactor

    International Nuclear Information System (INIS)

    Smith, R.D.; Regazzoni, Pierre.

    1981-01-01

    This invention concerns a gamma ray thermometer for nuclear reactors, fitted with a thermal bridge for use as a centring device. In accordance with the invention, an elastic device fills all the annular space between the gamma thermometer and the orifice through which the thermometer is introduced. This elastic device has the two-fold role of providing a thermal bridge at the gamma thermometer location suitable as a heat well, and of acting as a device for centring the thermometer in the orifice into which it has been introduced [fr

  7. Quasielastic neutron scattering facility at Dhruva reactor

    International Nuclear Information System (INIS)

    Mukhopadhyay, R.; Mitra, S.; Paranjpe, S.K.; Dasannacharya, B.A.

    2001-01-01

    Quasi-elastic neutron scattering is a powerful experimental tool for studying the various dynamical motions in solids and liquids. In this paper, we have described the salient features of the quasi-elastic neutron spectrometer in operation at Dhruva reactor at Trombay, India. The design criteria have been such as to maximise the throughput by various means like closer approach to the source, focusing a larger beam on to a sample, and Multi-Angle Reflecting X-tal mode of energy analysis. Some results of molecular motions from recently studied systems using this spectrometer are also reported

  8. CEDM Controller for a Linear Pulse Motor by using Pulse Width Modulation Method in Integral Reactor

    International Nuclear Information System (INIS)

    Lee, Joon-Koo; Keum, Jong-Yong; Park, Heui-Youn

    2007-01-01

    Integral Reactor SMART is under development at KAERI. The design characteristics of SMART are radically different from those employed in currently operating loop type PWR in Korea. The reliability and accuracy of Control Rod Drive Mechanism are very important to the reactor safety and the design of the Plant Protection System. The SMART CEDM designed for fine-step movement consists of a linear pulse motor, reed switch type sensor with top and bottom limit switches which also act as Control Element Assembly(CEA) Position indicator, The linear pulse motor is a four phase synchronous DC electric machine with inner stator and output stator in coolant medium inside a strong housing. The objective of this paper is to introduce and to explain the CEDM controller CEDM Controller is being developed with a new design concept and digital technology to reduce the Operating Error and improve the systems' reliability and availability. And Switched Mode Power Supply is also being developed with digital hardware technology. This paper involves the test details and result

  9. Pressure suppression facility for reactor container

    International Nuclear Information System (INIS)

    Fujii, Tadashi; Fukui, Toru; Kataoka, Yoshiyuki; Tominaga, Kenji.

    1993-01-01

    In a nuclear reactor comprising heat transfer surfaces from a pressure suppression pool at the inside to the outer circumferential pool at the outside, a means for supplying water from a water supply source at the outside of the container to the pools is disposed. Then, a heat transfer means is disposed between the pressure suppression chamber and the water cooling pool. The water supply means comprises a pressurization means for applying pressure to water of the water supply source and a water supply channel. Water is supplied into the pressure suppression pool and the outer circumferential pool to elevate the water level and extend the region of heat contact with the water cooling heat transfer means. In addition, since dynamic pressure is applied to the feedwater, for example, by pressurizing the water surface of the water supply source, water can be supplied without using dynamic equipments such as pumps. Then, since water-cooling heat transfer surface can be extended after occurrence of accident, enlargement of a reactor container and worsening of earthquake proofness can be avoided as much as possible, to improve function for suppressing the pressure in the container. Further, since water-cooling heat transfer region can be extended, the arrangement of the water source and the place to which water is supplied is made optional without considering the relative height therebetween, to improve earthquake proofness. (N.H.)

  10. 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)

  11. 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)

  12. Report of the Panel on Kinetics and Applications of Pulsed Research Reactors

    International Nuclear Information System (INIS)

    1966-03-01

    The question of the dynamic behaviour of a reactor subjected to a highly supercritical condition has had special interest for reactor physicists because of the reactor safety implications involved. The large amount of experimental and theoretical work done during the past dozen years or sc to understand fast transient behaviour and the inherent safety characteristics of reactors has not only helped to ease the concern of reactor designers about the consequences of a prompt critical excursion, but, by demonstrating the feasibility of operating certain types of reactors in a pulsed fashion has led to the development of an extremely useful research tool. Pulsed research reactors of a number of different kinds are in operation, while newer, higher performance systems are presently being designed and constructed. Such devices are being used more and more for research in physics, chemistry and reactor engineering, and with the advent of the newer machines, new research areas will become accessible. Because of the rapidly growing interest in the utilization of pulsed reactors for research, the IAEA convened a panel of experts in this field to review recent progress in the design and application of pulsed reactors to consider the problems of converting an existing pool type research reactor to a pulsing types and to consider future potentialities. The panel met in Vienna from 17 to 21 May 1965. This report of the panel summarizes the discussions

  13. Development of pulsed plate columns for fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Jenkins, J.A.; Logsdail, D.H.; Lyall, E.; Myers, P.E.; Partridge, B.A.

    1987-01-01

    The UK Atomic Energy Authority has undertaken a development programme on solvent extraction equipment for reprocessing fast reactor fuels. As part of this programme a solvent extraction pilot plant has been built at Harwell in which a variety of flowsheet conditions can be simulated using the system uranyl nitrate/nitric acid (UN/HNO 3 ) - 20% tri-n-butyl phosphate in odourless kerosene (TBP/OK). The main purpose of present pilot plant operations is to study the performance of pulsed plate columns, with the following specific objectives: to measure the volumetric throughput capacity of the columns, - to study the effect of scale-up of column diameter on U mass transfer performance, - to provide hydraulic and mass transfer data for a dynamic simulation model of pulsed column operation, - to develop and test instruments and ancillary equipment. This poster describes the pilot plant and is illustrated by experimental data, with particular reference to an external settler for controlling the removal of aqueous phase from columns operated with the aqueous phase dispersed

  14. Jordan Research and Training Reactor (JRTR) Utilization Facilities

    International Nuclear Information System (INIS)

    Xoubi, N.

    2013-01-01

    Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

  15. Nuclear piston engine and pulsed gaseous core reactor power systems

    International Nuclear Information System (INIS)

    Dugan, E.T.

    1976-01-01

    The investigated nuclear piston engines consist of a pulsed, gaseous core reactor enclosed by a moderating-reflecting cylinder and piston assembly and operate on a thermodynamic cycle similar to the internal combustion engine. The primary working fluid is a mixture of uranium hexafluoride, UF 6 , and helium, He, gases. Highly enriched UF 6 gas is the reactor fuel. The helium is added to enhance the thermodynamic and heat transfer characteristics of the primary working fluid and also to provide a neutron flux flattening effect in the cylindrical core. Two and four-stroke engines have been studied in which a neutron source is the counterpart of the sparkplug in the internal combustion engine. The piston motions which have been investigated include pure simple harmonic, simple harmonic with dwell periods, and simple harmonic in combination with non-simple harmonic motion. The results of the conducted investigations indicate good performance potential for the nuclear piston engine with overall efficiencies of as high as 50 percent for nuclear piston engine power generating units of from 10 to 50 Mw(e) capacity. Larger plants can be conceptually designed by increasing the number of pistons, with the mechanical complexity and physical size as the probable limiting factors. The primary uses for such power systems would be for small mobile and fixed ground-based power generation (especially for peaking units for electrical utilities) and also for nautical propulsion and ship power

  16. Sea water take-up facility for cooling reactor auxiliary

    International Nuclear Information System (INIS)

    Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

    1997-01-01

    The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

  17. Abnormal Events for Reactor System and Facilities in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ho Young; Lee, B. H.; Lee, M.; Kang, I. H.; Lee, U. G.; Sin, H. C.; Park, C. Y.; Song, B. S.; Lee, S. H.; Han, J. S

    2006-12-15

    This report gathers abnormal events related to reactor system and facilities of HANARO that happened during its operation over 10 years since the first criticality on February 1995. The collected examples will be utilized to the HANARO's operators as a useful guide.

  18. The neutron beam facility at the Australian replacement research reactor

    International Nuclear Information System (INIS)

    Hunter, B.; Kennedy, S.

    1999-01-01

    Full text: The Australian federal government gave ANSTO final approval to build a research reactor to replace HIFAR on August 25th 1999. The replacement reactor is to be a multipurpose reactor with a thermal neutron flux of 3 x 10 14 n.cm -2 .s -1 and having improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The replacement reactor will commence operation in 2005 and will cater for Australian scientific, industrial and medical needs well into the 21st century. The scientific capabilities of the neutron beams at the replacement reactor are being developed in consultation with representatives from academia, industry and government research laboratories to provide a facility for condensed matter research in physics, chemistry, materials science, life sciences, engineering and earth sciences. Cold, thermal and hot neutron sources are to be installed, and neutron guides will be used to position most of the neutron beam instruments in a neutron guide hall outside the reactor confinement building. Eight instruments are planned for 2005, with a further three to be developed by 2010. A conceptual layout for the neutron beam facility is presented including the location of the planned suite of neutron beam instruments. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by an accredited reactor builder in a turnkey contract. Tenders have been called for December 1999, with selection of contractor planned by June 2000. The neutron beam instruments will be developed by ANSTO and other contracted organisations in consultation with the user community and interested overseas scientists. The facility will be based, as far as possible, around a neutron guide hall that is be served by three thermal and three cold neutron guides. Efficient transportation of thermal and cold neutrons to the guide hall requires the use of modern super

  19. Linear pulse motor type control element drive mechanism for the integral reactor

    International Nuclear Information System (INIS)

    Yu, J. Y.; Choi, S.; Kim, J. H.; Huh, H.; Park, K. B.

    2007-01-01

    The integral reactor SMART currently under development at Korea Atomic Energy Research Institute is designed with soluble boron free operation and use of nuclear heating for reactor startup. These design features require the Control Element Drive Mechanism (CEDM) for SMART to have fine-step movement capability as well as high reliability for the fine reactivity control. In this paper, design characteristics of a new concept CEDM driven by the Linear Pulse Motor (LPM) which meets the design requirements of the integral reactor SMART are introduced. The primary dimensions of the linear pulse motor are determined by the electro-magnetic analysis and the results are also presented. In parallel with the electro-magnetic analysis, the conceptual design of the CEDM is visualized and checked for interferences among parts by assembling three dimensional (3D) models on the computer. Prototype of LPM with double air-gaps for the CEDM sub-assemblies to lift 100 kg is designed, analysed, manufactured and tested to confirm the validity of the CEDM design concept. A converter and a test facility are manufactured to verify the dynamic performance of the LPM. The mover of the LPM is welded with ferromagnetic material and non-ferromagnetic material to get the magnetic flux path between inner stator and outer stator. The thrust forces of LPM predicted by analytic model have shown good agreement with experimental results from the prototype LPM. It is found that the LPM type CEDM has high force density and simple drive mechanism to reduce volume and satisfy the reactor operating circumstances with high pressure and temperature

  20. All Solid State Optical Pulse Shaper for the OMEGA Laser Fusion Facility

    International Nuclear Information System (INIS)

    Okishev, A.V.; Skeldon, M.D.; Keck, R.L.; Seka, W.

    2000-01-01

    OAK-B135 All Solid State Optical Pulse Shaper for the OMEGA Laser Fusion Facility. The authors have developed an all-solid-state, compact, computer-controlled, flexible optical pulse shaper for the OMEGA laser facility. This pulse shaper produces high bandwidth, temporally shaped laser pulses that meet OMEGA requirements. The design is a significant simplification over existing technology with improved performance capabilities

  1. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

    2012-01-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  2. Environmental assessment for the deactivation of the N Reactor facilities

    International Nuclear Information System (INIS)

    1995-05-01

    This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. As required by the National Environmental Policy Act of 1969 (NEPA), this EA complies with Title 40, Code of Federal Regulations (CFR), parts 1500--1508, ''Regulations for Implementing the Procedural Provisions of NEPA. '' It also implements the ''National Environmental Policy Act; Implementing Procedures and Guidelines'' (10 CFR 1021)

  3. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    International Nuclear Information System (INIS)

    Stephens, S.V.

    1976-09-01

    The objective of the document is to present in one volume an overview of the Argonne National Laboratory test facilities involved in the conduct of the national LMFBR research and development program. Existing facilities and those under construction or authorized as of September 1976 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. The volume is divided into two sections: Argonne-East and Argonne-West. Introductory material for each section includes site and facility maps. The profiles are arranged alphabetically by title according to their respective locations at Argonne-East or Argonne-West. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc., involved in the LMFBR program is appended

  4. Ventilation safety of facilities comprising nuclear reactors

    International Nuclear Information System (INIS)

    Guirlet, J.

    1982-01-01

    The reliability of the ventilation is one of the most important aspects in the prevention of the nuisances that a nuclear installation can provide, since the ventilation is located at the last barrier. A certain number of essential points have been recalled here. But it is necessary to bear in mind other requirements such as the limitation in the number of crossovers, the answers to be found should the system fail, the need to show that ventilation systems do not in themselves bring other nuisances such as noise, irradiation or contamination hazards, likelyhood of recycling the contamination, vibrations, fire. Finally, it is absolutely essential, right from the project stage, that the design ensures that very good accessibility, very easy dismantling and handling, as well as all the facilities needed to make sure of the initial and periodic tests, are guaranteed [fr

  5. Neutron beam facilities at the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Kim, S.

    2003-01-01

    The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

  6. Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

    International Nuclear Information System (INIS)

    Kawahara, Toshio; Matsushita, Tadashi

    1977-01-01

    The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

  7. A neutron radiography facility on the IRT-2000 reactor

    International Nuclear Information System (INIS)

    Khadduri, I.Y.

    1976-01-01

    A neutron radiography facility has been constructed on the thermal neutron channel of the IRT-2000 reactor. A collimated thermal neutron beam exposure area of 10 cm diameter is obtained with an L/D ratio of 48.8. The film used is cellulose nitrate coated with lithium tetraborate which is insensitive to gamma and beta radiation. Some pictures with good contrast and resolution have been obtained. Pictures of parts of an IRT-2000 reactor fuel pin have also been recorded. (orig) [de

  8. Research reactor facilities, recent developments at Imperial College, London

    International Nuclear Information System (INIS)

    Franklin, S.J.; Goddard, A.J.H.; O Connell, J.

    1998-01-01

    The 100 kW CONSORT pool-type reactor is now the only Research Reactor in the UK. Because of its strategic importance, Imperial College is continuing and accelerating a programme of refurbishment of the control system, and planning for a further fuel charge. These plans are described and the progress to date discussed. To this end, a description of the enhanced Safety Case being written is provided here and refueling plans discussed. The current range of facilities available is described, and future plans highlighted. (author)

  9. Low temperature irradiation facility at Kyoto University Reactor (KUR)

    International Nuclear Information System (INIS)

    Atobe, Kozo; Okada, Moritami; Yoshida, Hiroyuki; Kodaka, Hisao; Miyata, Kiyomi.

    1977-01-01

    A new refrigeration system has been substituted to the low temperature irradiation facility at KUR instead of the previous one, since April in 1975. The model 1204 CTi He liquifier was designed to be modified for the refrigerator with the capacity of 30 watts at 10 K. The refrigeration capacity of 38 watts at 10 K was defined using a special cryostat and transfer-tubes, and the lowest temperature of about 18 K was measured using the irradiation loop without reactor operation. The reconstructed facility enables us to hold the many specimens simultaneously in the sample chamber of the irradiation loop at about 25 K during reactor operation of 5 MW. The irradiation dose has been reached about 6.6 x 10 16 n sub(f)/cm 2 and 6.1 x 10 17 n sub(th)/cm 2 with the normal reactor operation cycle of up to 77 hours. The stable operation condition of the machine and the special safety system for the refrigeration system enable us to maintain easily the facility with a constant operation condition for such a long time irradiation. Many kinds of low temperature neutron irradiation experiments are carried out using the facility, which techniques are partially reported. (auth.)

  10. Improvements in detection system for pulse radiolysis facility

    CERN Document Server

    Rao, V N; Manimaran, P; Mishra, R K; Mohan, H; Mukherjee, T; Nadkarni, S A; Sapre, A V; Shinde, S J; Toley, M

    2002-01-01

    This report describes the improvements made in the detection system of the pulse radiolysis facility based on a 7 MeV Linear Electron Accelerator (LINAC) located in the Radiation Chemistry and Chemical Dynamics Division of Bhabha Atomic Research Centre. The facility was created in 1986 for kinetic studies of transient species whose absorption lies between 200 and 700 nm. The newly developed detection circuits consist of a silicon (Si) photodiode (PD) detector for the wavelength range 450-1100 nm and a germanium (Ge) photodiode detector for the wavelength range 900-1600 nm. With these photodiode-based detection set-up, kinetic experiments are now routinely carried out in the wavelength range 450-1600 nm. The performance of these circuits has been tested using standard chemical systems. The rise time has been found to be 150 ns. The photo-multiplier tube (PMT) bleeder circuit has been modified. A new DC back-off circuit has been built and installed in order to avoid droop at longer time scales. A steady baselin...

  11. Neutron beam facilities at the Australian Replacement Research Reactor

    International Nuclear Information System (INIS)

    Kennedy, Shane; Robinson, Robert; Hunter, Brett

    2001-01-01

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10 14 n/cm 2 /sec and a liquid D 2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  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. 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.

  14. UCN-VCN facility and experiments in Kyoto University Reactor

    International Nuclear Information System (INIS)

    Kawabata, Yuji; Okumura, Kiyoshi; Utsuro, Masahiko

    1993-01-01

    An ultracold and very cold neutron facility was installed in Kyoto University Reactor (KUR). The facility consists of a very cold neutron (VCN) guide tube, a VCN bender, a supermirror neutron turbine and experimental equipments with ultracold neutrons (UCN). The properties of each equipments are presented. UCN is generated by a supermirror neutron turbine combined with the cold neutron source operated with liquid deuterium, and the UCN output spectrum was measured by the time-of-flight method. A gravity analyzer for high resolution spectroscopy and a neutron bottle for decay experiments are now developing as the UCN research in KUR. (author)

  15. The reactor and cold neutron research facility at NIST

    Energy Technology Data Exchange (ETDEWEB)

    Prask, H J; Rowe, J M [Reactor Radiation Division, National Institute of Standards and Technology, Gaithersburg, MD (United States)

    1992-07-01

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D{sub 2}O/H{sub 2}O ice cold source. (author)

  16. The reactor and cold neutron research facility at NIST

    International Nuclear Information System (INIS)

    Prask, H.J.; Rowe, J.M.

    1992-01-01

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D 2 O/H 2 O ice cold source. (author)

  17. Diamond Ordinance Radiation Facility (DORF) reactor operating experiences

    International Nuclear Information System (INIS)

    Gieseler, Walter

    1970-01-01

    The Diamond Ordnance Radiation Facility Mark F Reactor is described and some of the problems encountered with its operation are discussed. In a period from reactor startup in September 1961 to June 1964, when the aluminum-clad core was changed to a stainless-steel clad core, a total of 30 fuel elements were removed from reactor service because of excessive growth. One leaking fuel element was detected during the lifetime of the aluminum- clad core. In June 1964, the core was changed to the stainless-steel-clad high hydride fuel elements. Since the installation of the stainless-steel-clad fuel element core, there has been a gradual decline of excess reactivity. Various theories were discussed as the cause but the investigations have resulted in no definitive conclusion that could account for the total reactivity loss

  18. Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

    International Nuclear Information System (INIS)

    Karpenko, V.N.

    1983-01-01

    The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 10 14 cm -3 .s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

  19. The proposed cold neutron irradiation facility at the Breazeale reactor

    International Nuclear Information System (INIS)

    Dimeo, R. M.; Sokol, P. E.; Carpenter, J. M.

    1997-01-01

    We discuss the design considerations of a Cold Neutron Irradiation Facility (CNIF) originally to have been installed at the Penn State Breazeale Reactor (PSBR). The goal of this project was to study the effects of radiation-induced damage to cryogenic moderators and, in particular, solid methane. This work evolved through the design stage undergoing a full safety analysis and received tentative approval from the PSBR Safeguards Committee but was discontinued due to budgetary constraints. (auth)

  20. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    Kohn, E.; Wright, M.G.

    1976-11-01

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  1. Training and research reactor facility longevity extension program

    International Nuclear Information System (INIS)

    Carriveau, G.W.

    1991-01-01

    Since 1943, over 550 training and research reactors have been in operation. According to statistics from the International Atomic Energy Agency, ∼325 training and research reactors are currently in service. This total includes a wide variety of designs covering a range of power and research capabilities located virtually around the world. A program has been established at General Atomics (GA) that is dedicated to the support of extended longevity of training and research reactor facilities. Aspects of this program include the following: (1) new instrumentation and control systems; (2) improved and upgraded nuclear monitoring and control channels; (3) facility testing, repair and upgrade services that include (a) pool or tank integrity, (b) cooling system, and (c) water purification system; (4) fuel element testing procedures and replacement; (5) control rod drive rebuilding and upgrades; (6) control and monitoring system calibration and repair service; (7) training services, including reactor operations, maintenance, instrumentation calibration, and repair; and (8) expanded or new uses such as neutron radiography and autoradiography, isotope production, nuclear medicine, activation analysis, and material properties modification

  2. Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    De, S.K.; Karhadkar, C.G.

    2017-01-01

    An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

  3. The application of SQUG to non-reactor facilities

    International Nuclear Information System (INIS)

    Hoskins, R.S.

    1993-01-01

    DOE Order 6430.1A mandates that facilities be designed and constructed to withstand Natural Phenomena Hazards in accordance with their hazard level. DOE has a program in progress to evaluate and then upgrade many of their existing medium and high hazard facilities where release of hazardous materials to the environment is a concern. This paper addresses a useful methodology which has been applied by SRS to evaluate and qualify equipment to withstand the ravenousness of earthquakes. The Seismic Qualification Utility Group was formed by a group of Electric Power Utilities whose Nuclear Power Plants predated the 10CFR50 Environmental design requirements to develop a methodology to evaluate and upgrade their operating plants against seismic events in answer to NRC generic letter USI-A46. SRS participated in this organization, since it operated reactors designed and constructed in the 1950's, and the application of the SQUG methodology was obvious. Nuclear Material Processing and Handling (NMPH) facilities utilize equipment similar to the nuclear industry, in fact, to industry in general. Consequently, it made sense to apply SQUG methodology to evaluate and qualify equipment in NMPH facilities against earthquakes. In order to utilize SQUG methodology, some changes are required since the goal of safe shutdown of a NMPH facility differs from a nuclear reactor, consequently, the generation of a Safe Shutdown Equipment List is modified by the requirements of each particular facility. Once the Safety Class Equipment List is developed denoting the qualification requirement for each piece of equipment, the SQUG walk downs can be performed as they are in commercial nuclear plants. SQUG methodology offers a cost effective approach for the seismic qualification of equipment at existing DOE NMPH facilities. SQUG cannot be applied to new or unique equipment since the experience database doesn't contain the needed information

  4. Radiation protection planning for decommissioning of research reactor facilities

    International Nuclear Information System (INIS)

    Jackson, Roger; Harman, Neil; Craig, David; Fecitt, Lorna; Lobach, Yuri; Gorlinskij, Juri; Kolyadin, Vyacheslav; Pavlenko, Vytali

    2008-01-01

    The MR reactor at the Russian Research Centre Kurchatov Institute (RRCKI), Moscow was a 50 MW multipurpose material testing and research reactor equipped with nine experimental loop facilities to test prototype fuel for various nuclear power reactors being developed. The reactor was shut down in 1993 and de-fuelled. The experimental loops are located in basement rooms around the reactor. The nature of the research into the characteristics of fuel design and coolant chemistry resulted in fission products and activation products in the test loop equipment. Decommissioning of the loops therefore presents a number of challenges. In addition the city of Moscow has expanded such that the RRC KI is now surrounded by housing which had to be taken into account in the radiological protection planning. This paper describes the techniques proposed to undertake the dismantling operations in order to minimise the radiation exposure to workers and members of the public. Estimates have been made of the worker doses which could be incurred during the dismantling process and the environmental impacts which could occur. These are demonstrated to be as low as reasonably achievable. The work was funded by the UK Department of Business Enterprise and Regulatory Reform (DBERR) (formerly the Department of Trade and Industry) under the Nuclear Safety Programme (NSP) set up to address nuclear safety issues in the Former Soviet Union. (author)

  5. Atlas Pulsed Power Facility for High Energy Density Physics Experiments

    International Nuclear Information System (INIS)

    Miller, R.B.; Ballard, E.O.; Barr, G.W.; Bowman, D.W.; Chochrane, J.C.; Davis, H.A.; Elizondo, J.M.; Gribble, R.F.; Griego, J.R.; Hicks, R.D.; Hinckley, W.B.; Hosack, K.W.; Nielsen, K.E.; Parker, J.V.; Parsons, M.O.; Rickets, R.L.; Salazar, H.R.; Sanchez, P.G.; Scudder, D.W.; Shapiro, C.; Thompson, M.C.; Trainor, R.J.; Valdez, G.A.; Vigil, B.N.; Watt, R.G.; Wysock, F.J.

    1999-01-01

    The Atlas facility, now under construction at Los Alamos National Laboratory (LANL), will provide a unique capability for performing high-energy-density experiments in support of weapon-physics and basic-research programs. It is intended to be an international user facility, providing opportunities for researchers from national laboratories and academic institutions around the world. Emphasizing institutions around the world. Emphasizing hydrodynamic experiments, Atlas will provide the capability for achieving steady shock pressures exceeding 10-Mbar in a volume of several cubic centimeters. In addition, the kinetic energy associated with solid liner implosion velocities exceeding 12 km/s is sufficient to drive dense, hydrodynamic targets into the ionized regime, permitting the study of complex issues associated with strongly-coupled plasmas. The primary element of Atlas is a 23-MJ capacitor bank, comprised of 96 separate Marx generators housed in 12 separate oil-filled tanks, surrounding a central target chamber. Each tank will house two, independently-removable maintenance units, with each maintenance unit consisting of four Marx modules. Each Marx module has four capacitors that can each be charged to a maximum of 60 kilovolts. When railgap switches are triggered, the marx modules erect to a maximum of 240 kV. The parallel discharge of these 96 Marx modules will deliver a 30-MA current pulse with a 4-5-micros risetime to a cylindrical, imploding liner via 24 vertical, tri-plate, oil-insulated transmission lines. An experimental program for testing and certifying all Marx and transmission line components has been completed. A complete maintenance module and its associated transmission line (the First Article) are now under construction and testing. The current Atlas schedule calls for construction of the machine to be complete by August, 2000. Acceptance testing is scheduled to begin in November, 2000, leading to initial operations in January, 2001

  6. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

  7. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  8. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    International Nuclear Information System (INIS)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-01-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

  9. Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

    International Nuclear Information System (INIS)

    Chatt, A.; Holzbecher, J.

    1990-01-01

    SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

  10. Nature and characteristics of pulsing flow in trickle-bed reactors

    NARCIS (Netherlands)

    Boelhouwer, J.G.; Piepers, H.W.; Drinkenburg, A.A.H.

    2002-01-01

    Pulsing flow is well known for its advantages in terms of an increase in mass and heat transfer rates, complete catalyst wetting and a decrease in axial dispersion compared to trickle flow. The operation of a trickle-bed reactor in the pulsing flow regime is favorable in terms of a capacity increase

  11. Enlargement of the pulsing flow regime by periodic operation of a trickle-bed reactor.

    NARCIS (Netherlands)

    Boelhouwer, J.G.; Piepers, H.W.; Drinkenburg, A.A.H.

    1999-01-01

    Potential advantages of pulsing flow in trickle-bed reactors include capacity increase and elimination of hot spots through the enhanced mass and heat transfer rates. A disadvantage of naturally occurring pulsing flow is the necessity of relatively high gas and liquid flow rates, especially at

  12. On the controllability and run-away possibility of a totally free piston, pulsed compression reactor

    NARCIS (Netherlands)

    Roestenberg, T.; Glouchenkov, Maxim Joerjevisj; glushenkov, M.J.; Kronberg, Alexandre E.; van der Meer, Theodorus H.

    2010-01-01

    The pulsed compression reactor promises to be a compact, economical and energy efficient alternative to conventional chemical reactors. While its design and operation is similar to that of a free piston internal combustion engine, it does not benefit from any controllability through the load.

  13. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    International Nuclear Information System (INIS)

    Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P.

    1998-01-01

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

  14. 1996 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    1996-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 1996 at the Naval Reactors Facility (NRF) are presented in this report. The NRF is located on the Idaho National Engineering and Environmental Laboratory and contains three naval reactor prototypes and the Expended Core Facility, which examines developmental nuclear fuel material samples, spent naval fuel, and irradiated reactor plant components/materials. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  15. Simulation of emulsion copolymerization reactions in a continuous pulsed sieve-plate column reactor

    OpenAIRE

    C. Sayer; R. Giudici

    2004-01-01

    This work addressed the viability of using a pulsed sieve-plate column reactor to carry out continuous vinyl acetate/butyl acrylate emulsion copolymerization reactions. A rigorous mathematical model of emulsion copolymerization reactions in a tubular reactor with axial dispersion was used for this purpose. Operational conditions were defined to attain high monomer conversions at the reactor outlet in a relatively short residence time and, at the same time, produce a copolymer with a more homo...

  16. Simulation of emulsion copolymerization reactions in a continuous pulsed sieve-plate column reactor

    Directory of Open Access Journals (Sweden)

    Sayer C.

    2004-01-01

    Full Text Available This work addressed the viability of using a pulsed sieve-plate column reactor to carry out continuous vinyl acetate/butyl acrylate emulsion copolymerization reactions. A rigorous mathematical model of emulsion copolymerization reactions in a tubular reactor with axial dispersion was used for this purpose. Operational conditions were defined to attain high monomer conversions at the reactor outlet in a relatively short residence time and, at the same time, produce a copolymer with a more homogeneous composition.

  17. The 'MELUSINE' reactor at Grenoble, France, and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the MELUSINE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities and specialized irradiation devices (loops and capsules). The information is presented in the form of six information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities

  18. Preliminary design of the cooling system for a gas-cooled, high-fluence fast pulsed reactor (HFFPR)

    International Nuclear Information System (INIS)

    Monteith, H.C.

    1978-10-01

    The High-Fluence Fast Pulsed Reactor (HFFPR) is a research reactor concept currently being evaluated as a source for weapon effects experimentation and advanced reactor safety experiments. One of the designs under consideration is a gas-cooled design for testing large-scale weapon hardware or large bundles of full-length, fast reactor fuel pins. This report describes a conceptual cooling system design for such a reactor. The primary coolant would be helium and the secondary coolant would be water. The size of the helium-to-water heat exchanger and the water-to-water heat exchanger will be on the order of 0.9 metre (3 feet) in diameter and 3 metres (10 feet) in length. Analysis indicates that the entire cooling system will easily fit into the existing Sandia Engineering Reactor Facility (SERF) building. The alloy Incoloy 800H appears to be the best candidate for the tube material in the helium-to-water heat exchanger. Type 316 stainless steel has been recommended for the shell of this heat exchanger. Estimates place the cost of the helium-to-water heat exchanger at approximately $100,000, the water-to-water heat exchanger at approximately $25,000, and the helium pump at approximately $450,000. The overall cost of the cooling system will approach $2 million

  19. A system dynamics model for tritium cycle of pulsed fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Zuolong; Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Chen, Dehong, E-mail: dehong.chen@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

    2017-05-15

    As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

  20. A system dynamics model for tritium cycle of pulsed fusion reactor

    International Nuclear Information System (INIS)

    Zhu, Zuolong; Nie, Baojie; Chen, Dehong

    2017-01-01

    As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

  1. Personal neutron dosimetry at a research reactor facility

    International Nuclear Information System (INIS)

    Kamenopoulou, V.; Carinou, E.; Stamatelatos, I.E.

    2001-01-01

    Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve. (author)

  2. Gas cooled fast reactor background, facilities, industries and programmes

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1980-05-01

    This report was prepared at the request of the OECD-NEA Coordinating Group on Gas Cooled Fast Reactor Development and it represents a contribution (Vol.II) to the jointly sponsored Vol.I (GCFR Status Report). After a chapter on background with a brief description of the early studies and the activities in the various countries involved in the collaborative programme (Austria, Belgium, France, Germany, Japan, Sweden, Switzerland, United Kingdom and United States), the report describes the facilities available in those countries and at the Gas Breeder Reactor Association and the industrial capabilities relevant to the GCFR. Finally the programmes are described briefly with programme charts, conclusions and recommendations are given. (orig.) [de

  3. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

    The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

  4. HiRadMat: A high‐energy, pulsed beam, material irradiation facility

    CERN Multimedia

    Charitonidis, Nikolaos

    2016-01-01

    HiRadMat is a facility constructed in 2011, designed to provide high-intensity pulsed beams to an irradiation area where different material samples or accelerator components can be tested. The facility, located at the CERN SPS accelerator complex, uses a 440 GeV proton beam with a pulse length up to 7.2 μs and a maximum intensity up to 1E13 protons / pulse. The facility, a unique place for performing state-of-the art beam-to-material experiments, operates under transnational access and welcomes and financially supports, under certain conditions, experimental teams to perform their experiments.

  5. The neutron radiography facility at Tehran Research Reactor (TRR)

    International Nuclear Information System (INIS)

    Ali Pazirandeh

    2009-01-01

    Full text: Non-destructive testing in many fields of industry including detection of explosives, at the airports, testing for micro-cracks on airplane wings and turbine blades cracks is badly needed. Thermal neutron beam is one of preferable method to detect the micro-cracks, reveals the internal structure of components and explosives. The purpose of this paper is to present the neutron radiography facility at Tehran Research Reactor (TRR), Science and Technology Research Institute, and in particular to emphasize the industrial applications in wood industry, automobile engine inspection, minerals composition identification, turbine blade cracks detection. (author)

  6. Nuclear Security Management for Research Reactors and Related Facilities

    International Nuclear Information System (INIS)

    2016-03-01

    This publication provides a single source guidance to assist those responsible for the implementation of nuclear security measures at research reactors and associated facilities in developing and maintaining an effective and comprehensive programme covering all aspects of nuclear security on the site. It is based on national experience and practices as well as on publications in the field of nuclear management and security. The scope includes security operations, security processes, and security forces and their relationship with the State’s nuclear security regime. The guidance is provided for consideration by States, competent authorities and operators

  7. Remote maintenance in TOR fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

    Eymery, R.; Constant, M.; Malterre, G.

    1986-11-01

    The TOR facility which is undergoing commissioning tests has a capacity of 5 T. HM/year which is enough for reprocessing all the Phenix fuel, with an excess capacity which is to be used for other fast reactors fuels. It is the result of enlargement and renovation of the old Marcoule pilot facility. A good load factor is expected through the use of equipment with increased reliability and easy maintenance. TOR will also be used to test new equipment developed for the large breeder fuel reprocessing plant presently in the design stage. The latter objective is specifically important for the parts of the plant involving mechanical equipment which are located in a new building: TOR 1. High reliability and flexibility will be obtained in this building thanks to the attention given to the integrated remote handling system [fr

  8. Miniature fission chambers calibration in pulse mode: interlaboratory comparison at the. SCK·CEN BR1 and CEA CALIBAN reactors

    International Nuclear Information System (INIS)

    Lamirand, V.; Geslot, B.; Gregoire, G.; Garnier, D.; Breaud, S.; Mellier, F.; Di-Salvo, J.; Destouches, C.; Blaise, P.; Wagemans, J.; Borms, L.; Malambu, E.; Casoli, P.; Jacquet, X.; Rousseau, G.; Sauvecane, P.

    2013-06-01

    Miniature fission chambers are suited tools for instrumenting experimental reactors, allowing online and in-core neutron measurements of quantities such as fission rates or reactor power. A new set of such detectors was produced by CEA to be used during the next experimental program at the EOLE facility starting in 2013. Some of these detectors will be employed in pulse mode for absolute measurements, thus requiring calibration. The calibration factor is expressed in mass units and thus called 'effective mass'. A calibration campaign was conducted in December 2012 at the SCK.CEN BR1 facility within the framework of the scientific cooperation VEP (VENUS-EOLE-PROTEUS) between SCK.CEN, CEA and PSI. Two actions were conducted in order to improve the calibration method. First a new characterisation of the thermal flux cavity and the MARK3 neutron flux conversion device performed by SCK.CEN allowed using calculated effective cross sections for determining detectors effective masses. Dosimetry irradiations were performed in situ in order to determine the neutron flux level and provide link to the metrological standard. Secondly two fission chambers were also calibrated at the CEA CALIBAN reactor (fast neutron spectrum), using the same method so that the results can be compared with the results obtained at the SCK.CEN. In this paper the calibration method and recent improvements on uncertainty reduction are presented. The results and uncertainties obtained in the two reactors CALIBAN and BR1 are compared and discussed. (authors)

  9. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

  10. 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

  11. RIA testing capability of the transient reactor test facility

    International Nuclear Information System (INIS)

    Crawford, D.C.; Swanson, R.W.

    1999-01-01

    The advent of high-burnup fuel implementation in LWRs has generated international interest in high-burnup LWR fuel performance. Recent testing under simulated RIA conditions has demonstrated that certain fuel designs fail at peak fuel enthalpy values that are below existing regulatory criteria. Because many of these tests were performed with non-prototypically aggressive test conditions (i.e., with power pulse widths less than 10 msec FWHM and with non-protoypic coolant configurations), the results (although very informative) do not indisputably identify failure thresholds and fuel behavior. The capability of the TREAT facility to perform simulated RIA tests with prototypic test conditions is currently being evaluated by ANL personnel. TREAT was designed to accommodate test loops and vehicles installed for in-pile transient testing. During 40 years of TREAT operation and fuel testing and evaluation, experimenters have been able to demonstrate and determine the transient behavior of several types of fuel under a variety of test conditions. This experience led to an evolution of test methodology and techniques which can be employed to assess RIA behavior of LWR fuel. A pressurized water loop that will accommodate RIA testing of LWR and CANDU-type fuel has completed conceptual design. Preliminary calculations of transient characteristics and energy deposition into test rods during hypothetical TREAT RIA tests indicate that with the installation of a pressurized water loop, the facility is quite capable of performing prototypic RIA testing. Typical test scenarios indicate that a simulated RIA with a 72 msec FWHM pulse width and energy deposition of 1200 kJ/kg (290 cal/gm) is possible. Further control system enhancements would expand the capability to pulse widths as narrow as 40 msec. (author)

  12. Refueling the RPI reactor facility with low-enrichment fuel

    International Nuclear Information System (INIS)

    Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

    1985-01-01

    The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

  13. Facility Effluent Monitoring Plan for the N Reactor

    International Nuclear Information System (INIS)

    Watson, D.J.; Brendel, D.F.; Shields, K.D.

    1991-11-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP- 0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The primary purpose of the N Reactor Facility Effluent Monitoring Plan (FEMP), during standby, is to ensure that the radioactive effluents are properly monitored and evaluated for compliance with the applicable DOE orders and regulatory agencies at the federal, state, and local levels. A secondary purpose of the FEMP is to ensure that hazardous wastes are not released, in liquid effluents, to the environment even though the potential to do so is extremely low. The FEMP is to provide a monitoring system that collects representative samples in accordance with industry standards, performs analyses within stringent quality control (QC) requirements, and evaluates the data through the use of comparative analysis with the standards and acceptable environmental models

  14. On exposure of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1988

    International Nuclear Information System (INIS)

    1989-01-01

    The Law for Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1988 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the exposure of workers was below the permissible exposure dose in 1988 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of Japan Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  15. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    International Nuclear Information System (INIS)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO 2 fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject

  16. A seismic analysis of the driving system for the pulsed reactor

    International Nuclear Information System (INIS)

    Hu Yongtao; Fu Shixiang; Zeng Jianhua; Hong Jingfeng

    1991-01-01

    The driving system of the pulsed reactor contains control rods, pulsing o rod and sample rack. They are slender, and their drive function is required more strictly. First, a complete model which contains all driving system and reactor bridge is used. Then the substructure models are adopted. The results of calculation are compared with the experimental results. It shows that the analysis results are reliable and the substructure method is simple, available and utility. The seismic safety is evaluated by the results from response spectra method

  17. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles

  18. Performance of commercial off-the-shelf microelectromechanical systems sensors in a pulsed reactor environment

    Energy Technology Data Exchange (ETDEWEB)

    Hobert, Keith Edwin [Los Alamos National Laboratory; Heger, Arlen S [Los Alamos National Laboratory; Mccready, Steven S [Los Alamos National Laboratory

    2010-07-15

    Prompted by the unexpected failure of piezoresistive sensors in both an elevated gamma-ray environment and reactor core pulse tests, we initiated radiation testing of several MEMS piezoresistive accelerometers and pressure transducers to ascertain their radiation hardness. Some commercial off-the-shelf sensors are found to be viable options for use in a high-energy pulsed reactor, but others suffer severe degradation and even catastrophic failure. Although researchers are promoting the use of MEMS devices in radiation-harsh environment, we nevertheless find assurance testing necessary.

  19. Acoustic detection of boiling in the Sodium Loop Safety Facility in-reactor experiment P1

    International Nuclear Information System (INIS)

    Carey, W.M.; Anderson, T.T.; Bobis, J.P.

    1976-06-01

    Acoustic data were obtained from two high-temperature lithium niobate microphones on the loop background noise and transient pressure pulses during the Sodium Loop Safety Facility (SLSF) P1 in-reactor experiment. This experiment simulated an LMFBR loss-of-piping-integrity (LOPI) transient on a nineteen element, end-of-life, enriched-UO 2 fuel assembly. The microphones were exposed to liquid sodium at a distance 4.85 meters above the reactor core at temperatures between 315 0 and 590 0 C. The distance and location of the microphones in the P1 Test Train provided an attenuative transmission path which was undesirable for optimum acoustic detection of sodium boiling and fuel failure. The data gathered on the loop background noise was observed to be dominated by pump and electrical noise at frequencies below 1.5 KHz and appeared to be dominated by flow induced local turbulence noise at higher frequencies. During the period of time that the sodium in the fuel assembly was at its saturation temperature 943 0 C (1730 0 F), as indicated by the wire wrap thermocouples, several discrete pulses were observed with peak-to-peak pressure between 3.3 kPa and 7.9 kPa and center frequencies between 360 and 550 Hz. The pulses occurred at two separate gradually increasing repetition rates. These observations appear to be consistent with the result of an impulsive forcing function interacting with a band passed Helmholtz resonator. These data are consistent with the hypothesis that sodium boiling occurred in the P1 fuel assembly, resulting in the formation of individual voids that collapsed upon reaching the subcooled sodium. These data provide pertinent information regarding the feasibility of sodium boiling detection and may provide additional insight into the dynamics of the void behavior

  20. The selection of probabilistic safety assessment techniques for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    Vail, J.

    1992-01-01

    Historically, the probabilistic safety assessment (PSA) methodology of choice is the well known event tree/fault tree inductive technique. For reactor facilities is has stood the test of time. Some non-reactor nuclear facilities have found inductive methodologies difficult to apply. The stand-alone fault tree deductive technique has been used effectively to analyze risk in nuclear chemical processing facilities and waste handling facilities. The selection between the two choices suggest benefits from use of the deductive method for non-reactor facilities

  1. The 'SILOE' reactor at Grenoble, France and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the SILOE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  2. The DR 3 reactor at Risoe, Denmark and its associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DR 2 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of seven information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  3. The DIDO-reactor at Harwell, U.K. and ancillary hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DIDO reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  4. The 'OSIRIS' reactor at Saclay, France and available hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the OSIRIS reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  5. SANS facility at the Pitesti 14 MW Triga reactor

    International Nuclear Information System (INIS)

    Ionita, I.; Anghel, E.; Mincu, M.; Datcu, A.; Grabcev, B.; Todireanu, S.; Constantin, F.; Shvetsov, V.; Popescu, G.

    2006-01-01

    Full text of publication follows: At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. Examples of this type of applications are: 1) Phase separation and precipitates in material science, 2) Ultrafine grained materials (nano-crystals, ceramics), 3) Porous materials such as concretes and filter materials, 4) Conformation and entanglements of polymer-chains, 5) Aggregates of micelles in microemulsions, gels and colloids, 6) Radiation damage in steels and alloys. The need for the installation of a new SANS facility at the Triga Reactor of the Institute of Nuclear Researches in Pitesti, Romania become actual especially after the shutting down of the VVRS Reactor from Bucharest. A monochromatic neutron beam with 1.5 Angstrom ≤ λ ≤ 5 Angstrom is produced by a mechanical velocity selector with helical slots.The distance between sample and detectors plane is (5.2 m ). The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Q min = 0.005 Angstrom -1 while the maximal value is Q max = 0.5 Angstrom -1 . The relative error is ΔQ/Q min = 0.5. The cooperation partnership between advanced research centers and the smaller ones from developing countries could be fruitful. The formers act as mentors in solving specific problems. Such a partnership was established between INR Pitesti, Romania and JINR Dubna, Russia. The first step in this cooperation

  6. Design and characterisation of a pulsed neutron interrogation facility

    International Nuclear Information System (INIS)

    Favalli, A.; Pedersen, B.

    2007-01-01

    The Joint Research Centre recently obtained a license to operate a new experimental device intended for research in the field of nuclear safeguards. The research projects currently being planned for the new device includes mass determination of fissile materials in matrices and detection of contraband non-nuclear materials. The device incorporates a commercial pulsed neutron generator and a large graphite mantle surrounding the sample cavity. In this configuration, a relatively high thermal neutron flux with a long lifetime is achieved inside the sample cavity. By pulsing the neutron generator, a sample may be interrogated by a pure thermal neutron flux during repeated time periods. The paper reports on the design of the new device and the pulsed fast and thermal neutron source. The thermal neutron flux caused by the neutron generator and the graphite structure has been characterised by foil activation, fission chamber and 3 He proportional counter measurements. (authors)

  7. Naval Reactors Facility environmental monitoring report, calendar year 2001

    International Nuclear Information System (INIS)

    2002-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 2001 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U. S. Environmental Protection Agency and the U. S. Department of Energy

  8. 1997 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    1997-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 1997 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  9. Naval Reactors Facility Environmental Monitoring Report, Calendar Year 2003

    International Nuclear Information System (INIS)

    2003-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 2003 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency and the U.S. Department of Energy

  10. Naval Reactors Facility environmental monitoring report, calendar year 1999

    International Nuclear Information System (INIS)

    2000-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 1999 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE)

  11. 1993 environmental monitoring report for the naval reactors facility

    International Nuclear Information System (INIS)

    1994-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 1993 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  12. FFTF [Fast Flux Test Facility] reactor shutdown system reliability reevaluation

    International Nuclear Information System (INIS)

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations

  13. Nuclear blenders: blended learning from Rensselaer's Reactor Critical Facility

    International Nuclear Information System (INIS)

    Haley, T.C.

    2011-01-01

    Rensselaer's senior level undergraduate nuclear engineering course 'Critical Reactor Laboratory' is highly regarded and much loved. If you can get in, that is. But now it's a required course, nuclear engineering enrollment is up, and others are knocking on our door to get in. How might one offer such a unique course to the masses, without losing the whole point of a laboratory experience? This presentation looks at the costs and benefits of the transition to a 'blended learning' mode -- the merging of traditional, face-to-face instruction and web-based instruction as a solution. As part of the presentation, the course and the facility will be highlighted by short excepts from the 50 minute movie 'Everything You Always Wanted to Know about Neutron Chain Reactions (but were afraid to ask)'.

  14. Reactor cover gas monitoring at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Bechtold, R.A.; Holt, F.E.; Meadows, G.E.; Schenter, R.E.

    1986-09-01

    The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the US Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100-day operating cycle began in April 1982 and the eighth operating cycle was completed in July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification

  15. A new materials irradiation facility at the Kyoto university reactor

    International Nuclear Information System (INIS)

    Yoshiie, T.; Hayashi, Y.; Yanagita, S.; Xu, Q.; Satoh, Y.; Tsujimoto, H.; Kozuka, T.; Kamae, K.; Mishima, K.; Shiroya, S.; Kobayashi, K.; Utsuro, M.; Fujita, Y.

    2003-01-01

    A new materials irradiation facility with improved control capabilities has been installed at the Kyoto University Reactor (KUR). Several deficiencies of conventional fission neutron material irradiation systems have been corrected. The specimen temperature is controlled both by an electric heater and by the helium pressure in the irradiation tube without exposure to neutrons at temperatures different from the design test conditions. The neutron spectrum is varied by the irradiation position. Irradiation dose is changed by pulling the irradiation capsule up and down during irradiation. Several characteristics of the irradiation field were measured. The typical irradiation intensity is 9.4x10 12 n/cm 2 s (>0.1 MeV) and the irradiation temperature of specimens is controllable from 363 to 773 K with a precision of ±2 K

  16. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    International Nuclear Information System (INIS)

    Lewis, M.R.

    2000-01-01

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

  17. 1991 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    1991-01-01

    The results of the radiological and non-radiological environmental monitoring programs for 1991 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were within the guidelines established by state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or heath and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the EnVironmental Protection Agency (EPA) and the Department of Energy (DOE)

  18. Naval Reactors Facility environmental monitoring report, calendar year 2000

    International Nuclear Information System (INIS)

    2001-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 2000 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE)

  19. Reactor cover gas monitoring at the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, R A; Holt, F E; Meadows, G E; Schenter, R E [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    The Fast Flux Test Facility (FFTF) is a 400 megawatt (thermal) sodium cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the U. S. Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100 day operating cycle began in April 1982 and the eighth operating cycle was completed In July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification. A liquid argon Dewar system provides the large volume of inert gas required for operation of the FFTF. The gas is used as received and is not recycled. Low concentrations of krypton and xenon in the argon supply are essential to preclude interference with the gas tag system. Gas chromatography has been valuable for detection of inadvertent air in leakage during refueling operations. A temporary system is installed over the reactor during outages to prevent oxide formation in the sodium vapor traps upstream from the on line gas chromatograph. On line gas monitoring by gamma spectrometry and grab sampling with GTSTs has been successful for the identification of numerous radioactive gas releases from creep capsule experiments as well as 9 fuel pin ruptures. A redundant fission gas monitoring system has been installed to insure constant surveillance of the reactor cover gas.

  20. Program for studying fundamental interactions at the PIK reactor facilities

    International Nuclear Information System (INIS)

    Serebrov, A. P.; Vassiljev, A. V.; Varlamov, V. E.; Geltenbort, P.; Gridnev, K. A.; Dmitriev, S. P.; Dovator, N. A.; Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A.; Martemyanov, V. P.

    2016-01-01

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4′ channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4′ channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  1. Program for studying fundamental interactions at the PIK reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Vassiljev, A. V.; Varlamov, V. E. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation); Geltenbort, P. [Institut Laue-Langevin (France); Gridnev, K. A. [St. Petersburg State University (Russian Federation); Dmitriev, S. P.; Dovator, N. A. [Russian Academy of Sciences, Ioffe Physical-Technical Institute (Russian Federation); Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P. [National Research Center Kurchatov Institute (Russian Federation); and others

    2016-05-15

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4′ channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4′ channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  2. A Long-Pulse Spallation Source at Los Alamos: Facility description and preliminary neutronic performance for cold neutrons

    International Nuclear Information System (INIS)

    Russell, G.J.; Weinacht, D.J.; Pitcher, E.J.; Ferguson, P.D.

    1998-03-01

    The Los Alamos National Laboratory has discussed installing a new 1-MW spallation neutron target station in an existing building at the end of its 800-MeV proton linear accelerator. Because the accelerator provides pulses of protons each about 1 msec in duration, the new source would be a Long Pulse Spallation Source (LPSS). The facility would employ vertical extraction of moderators and reflectors, and horizontal extraction of the spallation target. An LPSS uses coupled moderators rather than decoupled ones. There are potential gains of about a factor of 6 to 7 in the time-averaged neutron brightness for cold-neutron production from a coupled liquid H 2 moderator compared to a decoupled one. However, these gains come at the expense of putting ''tails'' on the neutron pulses. The particulars of the neutron pulses from a moderator (e.g., energy-dependent rise times, peak intensities, pulse widths, and decay constant(s) of the tails) are crucial parameters for designing instruments and estimating their performance at an LPSS. Tungsten is the reference target material. Inconel 718 is the reference target canister and proton beam window material, with Al-6061 being the choice for the liquid H 2 moderator canister and vacuum container. A 1-MW LPSS would have world-class neutronic performance. The authors describe the proposed Los Alamos LPSS facility, and show that, for cold neutrons, the calculated time-averaged neutronic performance of a liquid H 2 moderator at the 1-MW LPSS is equivalent to about 1/4th the calculated neutronic performance of the best liquid D 2 moderator at the Institute Laue-Langevin reactor. They show that the time-averaged moderator neutronic brightness increases as the size of the moderator gets smaller

  3. On exposure management of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1993

    International Nuclear Information System (INIS)

    1994-01-01

    The Law of Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1993 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the the exposure of workers was below the permissible exposure dose in 1993 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of JAERI and PNC. (J.P.N.)

  4. Estimation of power feedback parameters of pulse reactor IBR-2M on transients

    International Nuclear Information System (INIS)

    Pepyolyshev, Yu.N.; Popov, A.K.

    2013-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) on a model of the reactor dynamics by mathematical treatment of two registered transients are estimated. Frequency characteristics and the pulse transient characteristics corresponding to these PFB parameters are calculated. PFB parameters received thus can be considered as their express tentative estimation as real measurements in this case occupy no more than 30 minutes. Total PFB is negative at 1 and 2 MW. At the received estimations of PFB parameters in a self-regulation mode it is possible to consider the stability margins of the IBR-2M reactor satisfactory

  5. Upgrade of the experimental facilities of the ORPHEE reactor

    International Nuclear Information System (INIS)

    Farnoux, B.; Breant, P.

    1993-01-01

    At the time of the design, the ORPHEE reactor has been equipped with a set of up-to-date experimental facilities such as nine tangential and horizontal beam holes, one hot source, two hydrogen cold sources and six neutron guides. After more than ten years of operations, all the neutron beams are now used by about twenty five spectrometers. A modernisation program is under progress with a two fold aim: upgrade of the existing facilities and creation of new beams. Some details of the six following points will be described: 1) replacement of the flat cold source cell by an hollow cylinder in order first to increase the cold neutron flux and secondly to facilitate the extraction of new cold neutron beams. 2) replacement of the old neutron guide elements coated with natural nickel by new elements with isotopic nickel or super mirror coating. 3) modification of the curvature of some existing neutron guides in order to increase the wavelength band transmission. 4) creation of new cold neutron beams by installation of benders on the existing neutron guides. 5) design of new cold neutron guides and a new guide hall. 6) design of a thermal neutron guide. The two last points will made extensive use of super mirrors allowed by new technical developments done at the Laboratoire LEON BRILLOUIN in connection with industry. (author)

  6. State of the art of nuclear facilities with organic cooled reactors

    International Nuclear Information System (INIS)

    Brede, O.

    1984-01-01

    USA, Canadian, and USSR activities aimed at developing nuclear facilities with organic cooled reactors are summarized. The facilities OMRE, PNPF, WR-1, and ARBUS are described, discussing in particular the problems of the chemistry of organic coolants. Finally, problems of further development and prospects of the application of organic cooled reactors are briefly outlined. (author)

  7. Improvement of pulsing operation performance in the Nuclear Safety Research Reactor (NSRR)

    International Nuclear Information System (INIS)

    Kobayasi, S.; Ishijima, K.; Tanzawa, S.; Fujishiro, T.; Horiki, O.

    1990-01-01

    The Nuclear Safety Research Reactor (NSRR) is one of the TRIGA-type research reactors widely used in the world, and has mainly been used for studying reactor fuel behaviour during postulated reactivity-initiated accidents (RIAs). Its limited pulsing operation capability, however, could produce only a power burst from low power level simulating an RIA event from essentially zero power level. A computerized automatic reactor control system was developed and installed in the NSRR to simulate a wide range of abnormal events in nuclear power plants. This digitalized reactor control system requires no manipulation of the control rods by reactor operators during the course of the pulsing operation. Using this fully automated operation system, a variety of power transients such as power ramping, power bursts from high power level, and so on were made possible with excellent stability and safety. The present modification work in the NSRR and its fruitful results indicate new possibilities in the utilization of the TRIGA type research reactor

  8. Analysis of Opportunity to Create Self-Regulating Reactor Facility of Extra-Low Power

    International Nuclear Information System (INIS)

    Kazansky, Y.A.; Levtchenko, V.A.; Yuriev, Y.S.

    2002-01-01

    This paper deals with fundamental possibilities (economy, safety, self-regulation) of creating an extra-low power reactor facility for heat supply. It contains the results of calculations for thermal and fast neutron reactors. The concept of this type of a reactor had been developed by the contributors earlier

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. Recent Development of Radioanalytical Methods at the IBR-2 Pulsed Fast Reactor

    International Nuclear Information System (INIS)

    Nazarov, V.M.; Peresedov, V.F.

    1994-01-01

    Experience in the application of radioanalytical methods, including NAA, at the IBR-2 pulsed fast reactor is reviewed. Details of the instruments dedicated to neutron activation analysis and radiography studies are reported. Applications of resonance neutrons to environmental monitoring and to the investigation of high-purity materials, are examplified. 15 refs. 9 figs., 9 tabs

  11. 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

  12. Conceptual design of a mirror reactor for a fusion engineering research facility (FERF)

    International Nuclear Information System (INIS)

    Batzer, T.H.; Burleigh, R.C.; Carlson, G.A.; Dexter, W.L.; Hamilton, G.W.; Harvey, A.R.; Hickman, R.G.; Hoffman, M.A.; Hooper, E.B. Jr.; Moir, R.W.; Nelson, R.L.; Pittenger, L.C.; Smith, B.H.; Taylor, C.E.; Werner, R.W.; Wilcox, T.P.

    1975-01-01

    A conceptual design is presented for a small mirror fusion reactor for a Fusion Engineering Research Facility (FERF). The reactor produces 3.4 MW of fusion power and a useful neutron flux of about 10 14 n.cm -2 .s -1 . Superconducting ''yin-yang'' coils are used, and the plasma is sustained by injection of energetic neutral D 0 and T 0 . Conceptual layouts are given for the reactor, its major components, and supporting facilities. (author)

  13. A pulse radiolysis based singlet oxygen luminescence facility

    International Nuclear Information System (INIS)

    Gorman, A.A.; Hamblett, I.; Land, E.J.

    1989-01-01

    In this paper the authors report the first successful time-resolved detection of singlet oxygen, O 2 ( 1 Δ g ), luminescence using the pulse radiolysis technique. The use of this technique (a) to produce high concentrations of solute (S) triplet states in aerated benzene (B) via a combination of channels 1-4 and (b) to subsequently form O 2 ( 1 Δ g ) via channel 5 has already been described. The method complements direct pulsed laser excitation of S in that formation of 3 S*, and therefore of O 2 ( 1 Δ g ), is still efficient in those instances where intersystem crossing (channel 4) is unimportant. In the latter situation a laser-based experiment would require an additional light-absorbing sensitizer which could subsequently transfer triplet energy to high concentrations of S. Such experiments, certainly of a quantitative nature, are usually doomed to failure because of competitive light absorption problems. No such problems exist with pulse radiolysis, and the high available triplet energy of 3 B* (84 kcal mol -1 ) ensures that virtually any solute of interest, in the O 2 ( 1 Δ g ) context, will be efficiently promoted to its triplet state

  14. The BNCT facility at the HFR Petten: Quality assurance for reactor facilities in clinical trials

    International Nuclear Information System (INIS)

    Moss, R.; Watkins, P.; Vroegindeweij, C.; Stecher-Rasmussen, F.; Huiskamp, R.; Ravensberg, K.; Appelman, K.; Sauerwein, W.; Hideghety, K.; Gabel, D.

    2001-01-01

    The first clinical trial in Europe of Boron Neutron Capture Therapy (BNCT) for the treatment of glioblastoma was opened in July 1997. The trial is a Phase I study with the principal aim to establish the maximum tolerated radiation dose and the dose limiting toxicity under defined conditions. It is the first time that a clinical application could be realised on a completely multi-national scale. The treatment takes place at the High Flux Reactor (HFR) in Petten, the Netherlands, is operated by an international team of experts under the leadership of a German radiotherapist, and treats patients coming from different European countries. It has therefore been necessary to create a very specialised organisation and contractual structure with the support of administrations from different countries, who had to find and adapt solutions within existing laws that had never foreseen such a situation. Furthermore, the treatment does not take place in an hospital environment and even more so, the facility is at a nuclear research reactor. Hence, special efforts were made on quality assurance, in order that the set-up at the facility and the personnel involved complied, as closely as possible, with similar practices in conventional radiotherapy departments. (author)

  15. An independent safety assessment of Department of Energy nuclear reactor facilities: Safety overview and management function

    International Nuclear Information System (INIS)

    Booth, M.; Brodsky, R.S.; Frankhouser, W.L.

    1981-02-01

    The Under Secretary of Energy established the Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee in October, 1979, in the aftermath of the Three Mile Island (TMI) nuclear accident, to assess the adequacy of training of personnel at DOE nuclear facilities. Subsequently, in February, 1980, the charge to this Committee was modified to assess all implications of the Kemeny Commission report on TMI with regard to DOE nuclear reactors, excluding those in the Division of Naval Reactors. The modified charge was also limited, for the time being, to reactor facilities instead of all nuclear facilities. This report describes the portion of the revised assessment activities that was assigned to the Assessment Support Team

  16. Concerning control of radiation exposure to workers in nuclear reactor facilities for testing and nuclear reactor facilities in research and development phase (fiscal 1987)

    International Nuclear Information System (INIS)

    1988-01-01

    A nuclear reactor operator is required by the Nuclear Reactor Control Law to ensure that the radiation dose to workers engaged in the operations of his nuclear reactor is controlled below the permissible exposure doses that are specified in notifications issued based on the Law. The present note briefly summarizes the data given in the Reports on Radiation Control, which have been submitted according to the Nuclear Reactor Control Law by the operators of nuclear reactor facilities for testing and those in the research and development phase, and the Reports on Control of Radiation Exposure to Workers submitted in accordance with the applicable administrative notices. According to these reports, the measured exposure to workers in 1987 were below the above-mentioned permissible exposure doses in all these nuclear facilities. The 1986 and 1987 measurements of radiation exposure dose to workers in nuclear reactor facilities for testing are tabulated. The measurements cover dose distribution among the facilities' personnel and workers of contractors. They also cover the total exposure dose for all workers in each of four plants operated under the Japan Atomic Energy Research Institute and the Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  17. Pulsed power supply and coaxial reactor applied to E. coli elimination in water by pulsed dielectric barrier discharge

    Energy Technology Data Exchange (ETDEWEB)

    Quiroz V, V. E.; Lopez C, R.; Rodriguez M, B. G.; Pena E, R.; Mercado C, A.; Valencia A, R.; Hernandez A, A. N.; Barocio, S. R.; Munoz C, A. E. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); De la Piedad B, A., E-mail: regulo.lopez@inin.gob.mx [Instituto Tecnologico de Toluca, Av. Tecnologico s/n, Ex-Rancho La Virgen, 52140 Metepec, Estado de Mexico (Mexico)

    2013-07-01

    The design and instrumentation intended for ATTC8739 Escherichia coli (E. coli) bacteria elimination in water, based on non thermal plasma generation at room pressure have been carried out by means of dielectric pulsed discharges. The latter have been produced by a power supply capable of providing voltages up to the order of 45 kV, 1-500 {mu}s pulse widths and variable frequencies between 100 Hz to 2000 Hz. This supply feeds a coaxial discharge reactor of the simple dielectric barrier type. The adequate operation of the system has been tested with the elimination of E. coli at 10{sup 4} and 10{sup 6} bacteria/ml concentrations, leading to reductions up to 85.3% and 95.1%, respectively, during the first 30 min of treatment. (Author)

  18. Description of the RA-3 research reactor as a model facility

    International Nuclear Information System (INIS)

    Vicens, Hugo E.; Quintana, Jorge A.

    2001-01-01

    The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

  19. Conversion of methane to hydrogen by a pulsed plasma reactor

    International Nuclear Information System (INIS)

    Ghorbanzadeh, A. M.; Matin, N.; Parandvar, M. R.; Rasouli, C.; Mazouchi, A. M.

    2003-01-01

    A pulsed atmospheric glow discharge, employing corona as a preionization, was used to convert methane to hydrogen and higher hydrocarbons. The experimental results showed that the overall conversion and specific energy, defined as energy needed to dissociate one mole methane, was mainly dependent on E/P, banking capacitance, repetition rate and flow rate. The dependence on E/P, especially, is more pronounced. The minimum specific energy was less than 1 MJ and it is expected that it could be further lowered by choosing higher E/P, lower banking capacitance and introducing an oxidizer to enhance the conversion efficiency

  20. Pulsed neutron source and instruments at neutron facility

    Energy Technology Data Exchange (ETDEWEB)

    Teshigawara, Makoto; Aizawa, Kazuya; Suzuki, Jun-ichi; Morii, Yukio; Watanabe, Noboru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    We report the results of design studies on the optimal target shape, target - moderator coupling, optimal layout of moderators, and neutron instruments for a next generation pulsed spallation source in JAERI. The source utilizes a projected high-intensity proton accelerator (linac: 1.5 GeV, {approx}8 MW in total beam power, compressor ring: {approx}5 MW). We discuss the target neutronics, moderators and their layout. The sources is designed to have at least 30 beam lines equipped with more than 40 instruments, which are selected tentatively to the present knowledge. (author)

  1. Syrinx - a research program for the pulsed power radiation facility

    International Nuclear Information System (INIS)

    Etlicher, B.; Chuvatin, A.S.; Choi, P.

    1996-01-01

    Syrinx is a targeted research program with the objective to study, through practical examples, the fundamentals necessary to define the details of all parts which will be required for a new powerful plasma radiation source. The current level of activities of Syrinx is in the design and construction of a multi-megajoule class IES based pulsed power driver which will use long conduction Plasma Opening Switch technology. The present paper reviews mainly the basic experimental research of the POS a nd Z-pinch accomplished in the framework of Syrinx project. This work has a unique international level of participation, from conceptual designs to particular investigations. (author). 9 figs., 17 refs

  2. Simulation of pulsed accidental energy release in a reactor core

    International Nuclear Information System (INIS)

    Ryshanskii, V.A.; Ivanov, A.G.; Uskov, A.A.

    1995-01-01

    At the present time the strength of the load-bearing members of VVER and fast reactors during a hypothetical accident is ordinarily investigated in model experiments [1]. A power burst during an accident is simulated by a nonnuclear exothermal reaction in water, which simulates the coolant and fills the model. The problem is to make the correct choice of the simulator of the accidental energy burst as an effective (i.e., sufficiently high working capacity) source of dangerous loads, corresponding to the conditions of an accident. What factors and parameters determine the energy release? The answers to these questions are contradictory

  3. Liner of a thermonuclear pulse THETA-pinch reactor

    International Nuclear Information System (INIS)

    Baranov, G.A.; Izotov, E.N.; Karasev, B.G.; Komin, A.V.; Krivosheev, M.V.; Levashov, A.D.

    1975-01-01

    Some possible constructive solutions to the problem of fabrication of the theta-pinch reactor liner by the method of centrifugal casting in a casting mould are considered. A scheme for liner manufacturing is presented, which includes the following elements: 1) a casting mould of dielectric material presenting a hollow cylinder of 4 m in diam., 3 m in length and 12 t in weight, which rotates at 8 rps in the reactor chamber; 2) a system for heat protection of the casting mould; the volume heat of the mould is suggested to remove by gaseous helium flowing under pressure along axial cooling channels of 5 mm in diam.; the channels are evenly distributed throughout the thickness of the mould shell; 3) a system for preparation and supply of a liquid metal to the casting mould, the metal is being supplied into the casting mould from its both ends at a rate of 1.7 t of the melt per second; 4) a system for rotation of the mould, which comprises two gas turbines mounted on both ends of the mould and two main stop-radial slip supports with gas lubrication

  4. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    International Nuclear Information System (INIS)

    Hu Jian; Jiang Nan; Li Jie; Shang Kefeng; Lu Na; Wu Yan; Mizuno Akira

    2016-01-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. (paper)

  5. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    Science.gov (United States)

    Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

    2016-03-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

  6. Operational characteristics of the CALIBAN fast pulse reactor

    International Nuclear Information System (INIS)

    Cortella, J.; Reberol, R.; Vanel, M.

    1976-01-01

    CALIBAN is a FPR operated at CEA-Valduc Center since 1971. It has been designed as a fast neutron irradiation source and its environment is specific for this utilization. To date, it delivered more than 400 bursts and the fuel integrated about 5.10 19 fissions. The main characteristics are: - cylindirical core 113 kg U 235 - Mo 10% alloy - integrated dose in a burst in the central 2.5cm diam cavity:3.10 14 n.cm -2 - integrated dose in a burst outside of the core:5.10 13 n.cm -2 - pulse width:50μs A special effort was made in measuring the spectrometric and dosimetric neutron and gamma characteristics. Some results will be presented here. (auth.)

  7. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  8. Seismic response analyses for reactor facilities at Savannah River

    International Nuclear Information System (INIS)

    Miller, C.A.; Costantino, C.J.; Xu, J.

    1991-01-01

    The reactor facilities at the Savannah River Plant (SRP) were designed during the 1950's. The original seismic criteria defining the input ground motion was 0.1 G with UBC [uniform building code] provisions used to evaluate structural seismic loads. Later ground motion criteria have defined the free field seismic motion with a 0.2 G ZPA [free field acceleration] and various spectral shapes. The spectral shapes have included the Housner spectra, a site specific spectra, and the US NRC [Nuclear Regulatory Commission] Reg. Guide 1.60 shape. The development of these free field seismic criteria are discussed in the paper. The more recent seismic analyses have been of the following type: fixed base response spectra, frequency independent lumped parameter soil/structure interaction (SSI), frequency dependent lumped parameter SSI, and current state of the art analyses using computer codes such as SASSI. The results from these computations consist of structural loads and floor response spectra (used for piping and equipment qualification). These results are compared in the paper and the methods used to validate the results are discussed. 14 refs., 11 figs

  9. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  10. Long-pulse beamlines for the mirror fusion test facility

    International Nuclear Information System (INIS)

    Stone, R.R.; Goldner, A.I.; Poulsen, P.

    1984-01-01

    We have recently obtained test results indicating that a beam of pure full-energy deuterium particles can be delivered to the plasma targets in MFTF-B. We used a close-coupled separator magnet with the ion source to separate the impurities from the full-energy deuterium particles. Our completed studies show that the usual iron-core sweep magnet and shielding used in neutral beamlines can be eliminated and the gas flow out of the beamline decreased. This design also reduces beam losses. We will use smooth-bore OFHC tube arrays brazed to manifolds for the active heat transfer surfaces. Tests indicate that both burnout and life requirements are met by this design. In this paper, we present test results and discuss the MFTF-B long-pulse beamline configuration

  11. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Science.gov (United States)

    2012-05-03

    ... Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY: Nuclear Regulatory Commission... Commission (NRC or the Commission) has issued renewed Facility Operating License No. R- 112, held by Reed... License No. R-112 will expire 20 years from its date of issuance. The renewed facility operating license...

  12. An independent safety assessment of Department of Energy nuclear reactor facilities: Procedures, operations and maintenance

    International Nuclear Information System (INIS)

    Toto, G.; Lindgren, A.J.

    1981-02-01

    The 1979 accident at the Three Mile Island commercial nuclear power plant has led to a number of studies of nuclear reactors, in both the public and private sectors. One of these is that of the Department of Energy's (DOE) Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, which has outlined tasks for assessment of 13 reactors owned by DOE and operated by contractors. This report covers one of the tasks, the assessment of procedures, operations, and maintenance at the DOE reactor facilities, based on a review of actual documents used at the reactor sites

  13. A pulsed neutron facility for condensed matter research

    International Nuclear Information System (INIS)

    Hobbis, L.C.W.; Rees, G.H.; Stirling, G.C.

    1977-06-01

    The scientific and technical basis of the project is presented, as follows: broad synopsis of the proposal for a spallation neutron facility; description of neutron scattering and current work in the UK; scientific applications of the Spallation Neutron Source; discussion of various types of neutron sources; outline description of the SNS and its neutron performance parameters; appendix dealing in more detail with utilization (solid state physics, fluids and amorphous solids, structure determination, molecular and biological sciences); appendix dealing in more detail with the project design (800 MeV synchrotron, target station, shielding, radioactivity and radiation damage, utilization, overall programme). (U.K.)

  14. A multichannel pulse acquisition system for reactor dosimetry data

    International Nuclear Information System (INIS)

    Talpalariu, C.; Talpalariu, J.; Matei, C.

    1999-01-01

    Simultaneous measurements of many dosimetry parameters require a complex instrumentation equipment, computers and interfaces. For a very large frequency range (10 -3 to 10 6 Hz) scale and mode selection (period or frequency), a big problem in multichannel pulse measurement is that of dead time, precision and response time. The dead time for normal scale selection and for data reading or writing for every channel can be as long as the active measuring time and response time for very large frequency variation can be very long, too. To solve this problem we have designed for a simultaneous 40 channel measurement, a pulse counter sampling system and an expert operating system. Based on a 486 PC and a 10 channel Timer/Counter Card, the hardware performance of the system was improved by an expert program for early rate change detection and rate prediction. The rate value was determined from optimizing between current value, medium and long time values and shorter response time for transient signals. Significant features and advantages of the system are the following: a marked reduction in panel complexity, as many of the indicators and controls can be replaced by an interactive CRT keyboard, a reduction in the instrumentation complexity, failure detection and diagnosis, system performance monitoring, intelligent alarm handling. The system was designed from high accuracy measurements on 40 simultaneous channels fed from field radiation detectors like ionizing chambers, fission chambers and photomultipliers.. The operating system is using an auto-organizing data memory for both computing the current value and for long-term management of data, so that only the status and significant values of the input are recorded. Consequently, the algorithms of decision, search and data processing are simplified and limited to the necessary memory, although enough memory is preserved for accurate representation of the dosimetry curves. The utilization of an inferential algorithm for the

  15. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    International Nuclear Information System (INIS)

    Tomberlin, T.A.

    2002-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed

  16. Energy balance and efficiency of power stations with a pulsed Tokamak reactor

    International Nuclear Information System (INIS)

    Davenport, P.A.; Mitchell, J.T.D.; Darvas, J.; Foerster, S.; Sack, B.

    1976-06-01

    The energy balance of a fusion power station based on the TOKAMAK concept is examined with the aid of a model comprising three distinct elements: the reactor, the energy converter and the reactor operation equipment. The efficiency of each element is expressed in terms of the various energy flows and the product of these efficiencies gives the net station efficiency. The analysis takes account of pulsed operation and has general applicability. Numerical values for the net station efficiency are derived from detailed estimates of the energy flows for a TOKAMAK reactor and its auxiliary equipment operating with advanced energy converters. The derivation of these estimates is given in eleven appendices. The calculated station efficiencies span ranges similar to those quoted for the current generation of fission reactors, though lower than those predicted for HTGR and LMFBR stations. Credible parameter domains for pulsed TOKAMAK operation are firmly delineated and factors inimical to improved performance are indicated. It is concluded that the net thermal efficiency of a TOKAMAK reactor power station based on present designs and using advanced thermal converters will be approximately 0.3 and is unlikely to exceed 0.33. (orig.) [de

  17. Design of the coolant system for the Large Coil Test Facility pulse coils

    International Nuclear Information System (INIS)

    Bridgman, C.; Ryan, T.L.

    1983-01-01

    The pulse coils will be a part of the Large Coil Test Facility in Oak Ridge, Tennessee, which is designed to test six large tokamak-type superconducting coils. The pulse coil set consists of two resistive coaxial solenoid coils, mounted so that their magnetic axis is perpendicular to the toroidal field lines of the test coil. The pulse coils provide transient vertical fields at test coil locations to simulate the pulsed vertical fields present in tokamak devices. The pulse coils are designed to be pulsed for 30 s every 150 s, which results in a Joule heating of 116 kW per coil. In order to provide this capability, the pulse coil coolant system is required to deliver 6.3 L/s (100 gpm) of subcooled liquid nitrogen at 10-atm absolute pressure. The coolant system can also cool down each pulse coil from room temperature to liquid nitrogen temperature. This paper provides details of the pumping and heat exchange equipment designed for the coolant system and of the associated instrumentation and controls

  18. 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

  19. 7-MeV electron LINAC based pulse radiolysis facility at RPCD, BARC

    International Nuclear Information System (INIS)

    Naik, C.B.; Nadkarni, S.A.; Toley, M.A.; Shinde, S.J.; Naik, P.D.

    2017-01-01

    7-MeV electron LINAC based pulse radiolysis facility is operational in Chemistry Group of BARC since 1986. The Accelerator is housed in B-132 room in basement of Modular Labs. BARC Accelerator was procured from Radiation Dynamics Inc. UK and its detection system was indigenously developed

  20. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  1. An optimization study of peak thermal neutron flux in moderators of advanced repetitive pulse reactors

    International Nuclear Information System (INIS)

    Asaoka, Takumi; Watanabe, N.

    1976-01-01

    In achieving a high peak thermal neutron flux in hydrogenous moderators installed in repetitive pulse reactors, the core-moderator arrangement can play as much an important role as the moderator design itself. However, the effect of the former has not been adequately emphasized to date, while a rather extensive study has been made on the latter. The present study concerns with a core-moderator system parameter optimization for a repetitive accelerator pulsed fast reactor. The results have shown that small differences in the arrangement resulting from the optimizations of various parameters are significant and the effects can be summed up to give an increase in the peak thermal flux by a factor of about two. (auth.)

  2. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

    1986-01-01

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

  3. The LNCMP: a pulsed-field user-facility in Toulouse

    International Nuclear Information System (INIS)

    2004-01-01

    Since summer 2002 the 'Laboratoire National des Champs Magnetiques Pulses' is operated as an international user facility providing access to pulsed magnetic fields up to 76 T. The laboratory disposes of 10 magnet stations equipped with long-pulse magnets operating in the 35-60 T range. A short-pulse system reaching 76 T was recently installed in the framework of the European ARMS-project. The experimental infrastructure includes various low-temperature systems ranging from ordinary flow-type cryostats to dilution refrigerators reaching 50 mK. A system permitting transport measurements under hydrostatic pressure and at low temperatures has recently been successfully tested. Experimental techniques include magnetization, transport, luminescence, IR-spectroscopy and the magneto-optical Kerr-effect. The LNCMP pursues an extensive in-house research program focusing on technological (reinforced wire, coil design, coil aging) as well as scientific aspects (magnetism, strongly correlated fermions, disordered systems and others)

  4. 50 mm Diameter digital DC/pulse neutron generator for subcritical reactor test

    International Nuclear Information System (INIS)

    Li Gang; Zhang Zhongshuai; Chi Qian; Liu Linmao

    2012-01-01

    A 50 mm diameter digital DC/pulse neutron generator was developed with 25 mm ceramic drive-in target neutron tube. It was applied in the subcritical reactor test of China Institute of Atomic Energy (CIAE). The generator can produce neutron in three modes: DC, pulse and multiple pulse. The maximum neutron yield of the generator is 1 × 10 8 n/s, while the maximum pulse frequency is 10 kHz, and the minimum pulse width is 10 μs. As a remote controlled generator, it is small in volume, easy to be connected and controlled. The tested results indicate that penning ion source has the feature of delay time in glow discharge, and it is easier for glow discharge to happen when switching the DC voltage of penning ion source into pulse. According to these two characteristics, the generator has been modified. This improved generator can be used in many other areas including Prompt Gamma Neutron Activation Analysis (PGNAA), neutron testing and experiment.

  5. 50 mm Diameter digital DC/pulse neutron generator for subcritical reactor test

    Energy Technology Data Exchange (ETDEWEB)

    Li Gang; Zhang Zhongshuai [Northeast Normal University, Changchun 130024 (China); Chi Qian [Guang Hua College of Chang Chun University, Changchun 130117 (China); Liu Linmao, E-mail: ll888@nenu.edu.cn [Northeast Normal University, Changchun 130024 (China)

    2012-11-01

    A 50 mm diameter digital DC/pulse neutron generator was developed with 25 mm ceramic drive-in target neutron tube. It was applied in the subcritical reactor test of China Institute of Atomic Energy (CIAE). The generator can produce neutron in three modes: DC, pulse and multiple pulse. The maximum neutron yield of the generator is 1 Multiplication-Sign 10{sup 8} n/s, while the maximum pulse frequency is 10 kHz, and the minimum pulse width is 10 {mu}s. As a remote controlled generator, it is small in volume, easy to be connected and controlled. The tested results indicate that penning ion source has the feature of delay time in glow discharge, and it is easier for glow discharge to happen when switching the DC voltage of penning ion source into pulse. According to these two characteristics, the generator has been modified. This improved generator can be used in many other areas including Prompt Gamma Neutron Activation Analysis (PGNAA), neutron testing and experiment.

  6. Feasibility study to develop BNCT facility at the Indonesian research reactor

    International Nuclear Information System (INIS)

    Hastowo, H.

    2001-01-01

    A survey on the Indonesian research reactors and its supporting facilities has been done in order to check the possibility to install BNCT facility. Oncologists from several hospitals have been informing about the BNCT treatment for tumours and they give a positive response to support utilisation of the BNCT facility. Several aspects required to support the BNCT treatment have also been identified and related activities on that matter soon will be initiated. The interim result in our survey indicated that utilisation of the 30 MW Multipurpose reactor would not be possible from the technical point of view. Further study will be concentrated on the TRIGA reactor and an epithermal neutron beam facility at the thermal column of this reactor will be designed for further work. (author)

  7. Model study of an automatic controller of the IBR-2 pulsed reactor

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.

    2007-01-01

    For calculation of power transients in the IBR-2 reactor a special mathematical model of dynamics taking into account the discontinuous jump of reactivity by an automatic controller with the step motor is created. In the model the nonlinear dependence of the energy of power pulse on the reactivity and the influence of warming up of the reactor on the reactivity by means of introduction of a nonlinear feedback 'power-pulse energy - reactivity' are taken into account. With the help of the model the transients of relative deviation of power-pulse energy are calculated at various (random, mixed and regular) reactivity disturbances at the reactor mean power 1.475 MW. It is shown that to improve the quality of processes the choice of such regular values of parameters of the automatic controller is expedient, at which the least effect of smoothing of a signal acting on an automatic controller and the least speed of an automatic controller are provided, and the reduction of efficiency of one step of the automatic controller and introduction of a five-percent dead space are also expedient

  8. Design and initial performance of the Sandia Pulsed Reactor-III

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Estes, B.F.

    1976-01-01

    The Sandia Pulsed Reactor-III (SPR-III) is a new fast pulsed reactor which has recently undergone initial testing at Sandia Laboratories. SPR-III is a uranium-10 weight percent molybdenum fuel assembly with a 17.78 cm irradiation cavity similar in design to SPR-II which has been in operation since 1967. The basic SPR-III design utilizes the same split-core configuration which has been proven with SPR-II; however, SPR-III uses external reflectors for control and external bolts to hold the fuel plates together. The core consists of sixteen fuel plates with an inside diameter of 17.78 cm, an outside diameter of 29.72 cm, and a core height of 31.9 cm. The fuel mass is about 227 kg of fully enriched uranium-10 weight percent molybdenum alloy. SPR III has completed the initial series of startup tests which included the critical experiment, zero and low-power tests, and pulse testing. The reactor design and results from the initial testing program are described in this paper. A portion of the startup experiments with SPR-III have been completed and this paper discusses the more important aspects of the initial testing program

  9. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  10. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  11. Alternative water injection device to reactor equipment facility

    International Nuclear Information System (INIS)

    Yamashita, Masahiro.

    1995-01-01

    The device of the present invention injects water to the reactor and the reactor container continuously for a long period of time for preventing occurrence of a severe accident in a BWR type reactor and maintaining the integrity of the reactor container even if the accident should occur. Namely, diesel-driven pumps disposed near heat exchangers of a reactor after-heat removing system (RHR) are operated before the reactor is damaged by the after heat to cause reactor melting. A sucking valve disposed to a pump sucking pipeline connecting a secondary pipeline of the RHR heat exchanger and the diesel driving pump is opened. A discharge valve disposed to a pump discharge pipeline connecting a primary pipeline of the RHR heat exchanger and the diesel driving pump is opened. With such procedures, sea water is introduced from a sea water taking port through the top end of the secondary pipeline of the RHR heat exchanger and water is injected into the inside of the pressure vessel or the reactor container by way of the primary pipeline of the RHR heat exchanger. As a result, the reactor core is prevented from melting even upon occurrence of a severe accident. (I.S.)

  12. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  13. Application of the pulsed magnetic welding process to nuclear breeder reactor fuel pin end closures

    International Nuclear Information System (INIS)

    Brown, W.F.

    1984-01-01

    The pulsed magnetic welding process is a solid state welding process in which metallurgical bonding is effected by impacting metal or alloy parts against each other at high velocity by use of controlled high frequency, high intensity pulsed magnetic fields. This process is similar to the explosive welding process except that magnetic energy is used for impacting the parts together instead of using explosive energy. The pulsed magnetic welding (PMW) process is readily applied to the welding of cylindrical plugs to small diameter tubes. Although breeder reactor fuel pin design may vary in size, the application described here consisted of cladding tubes approximately 6.4 mm in diameter by 244 cm long with a wall thickness of 0.38 mm. After the cladding tubes are filled with fuel pellets and associated metal hardware, tapered end plugs are inserted into the end of the tubes and welded. A typical setup for PMW is described

  14. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    International Nuclear Information System (INIS)

    Sager, G.T.; Wong, C.P.C.; Kapich, D.D.; McDonald, C.F.; Schleicher, R.W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankie and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed

  15. Gamma compensated pulsed ionization chamber wide range neutron/reactor power measurement system

    International Nuclear Information System (INIS)

    Ellis, W.H.

    1975-01-01

    An improved method and system of pulsed mode operation of ionization chambers is described in which a single sensor system with gamma compensation is provided by sampling, squaring, automatic gate selector, and differential amplifier circuit means, employed in relation to chambers sensitized to neutron plus gamma and gamma only to subtract out the gamma component, wherein squaring functions circuits, a supplemental high performance pulse rate system, and operational and display mode selection and sampling gate circuits are utilized to provide automatic wide range linear measurement capability for neutron flux and reactor power. Neon is employed as an additive in the ionization chambers to provide independence of ionized gas kinetics temperature effects, and the pulsed mode of operation provide independence of high temperature insulator leakage effects. (auth)

  16. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

    1983-02-01

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  17. Treatment of Dye Wastewater by Using a Hybrid Gas/Liquid Pulsed Discharge Plasma Reactor

    International Nuclear Information System (INIS)

    Lu Na; Li Jie; Wu Yan; Masayuki, Sato

    2012-01-01

    A hybrid gas/liquid pulsed discharge plasma reactor using a porous ceramic tube is proposed for dye wastewater treatment. High voltage pulsed discharge plasma was generated in the gas phase and simultaneously the plasma channel was permeated through the tiny holes of the ceramic tube into the water phase accompanied by gas bubbles. The porous ceramic tube not only separated the gas phase and liquid phase but also offered an effective plasma spreading channel. The effects of the peak pulse voltage, additive gas varieties, gas bubbling rate, solution conductivity and TiO 2 addition were investigated. The results showed that this reactor was effective for dye wastewater treatment. The decoloration efficiency of Acid Orange II was enhanced with an increase in the power supplied. Under the studied conditions, 97% of Acid Orange II in aqueous solution was effectively decolored with additive oxygen gas, which was 51% higher than that with argon gas, and the increasing O 2 bubbling rate also benefited the decoloration of dye wastewater. Water conductivity had a small effect on the level of decoloration. Catalysis of TiO 2 could be induced by the pulsed discharge plasma and addition of TiO 2 aided the decoloration of Acid Orange II.

  18. Rapid response and wide range neutronic power measuring systems for fast pulsed reactors

    International Nuclear Information System (INIS)

    Sumita, Kenji; Iida, Toshiyuki; Wakayama, Naoaki.

    1976-01-01

    This paper summarizes our investigation on design principles of the rapid, stable and wide range neutronic power measuring system for fast pulsed reactors. The picoammeter, the logarithmic amplifier, the reactivity meter and the neutron current chamber are the items of investigation. In order to get a rapid response, the method of compensation for the stray capacitance of the feedback circuits and the capacitance of signal cables is applied to the picoammeter, the logarithmic amplifier and the reactivity meter with consideration for the stability margin of a whole detecting system. The response of an ionization current chamber and the method for compensating the ion component of the chamber output to get optimum responses high pass filters are investigated. Statistical fluctuations of the current chamber output are also considered in those works. The optimum thickness of the surrounding moderator of the neutron detector is also discussed from the viewpoint of the pulse shape deformation and the neutron sensitivity increase. The experimental results are reported, which were observed in the pulse operations of the one shot fast pulsed reactor ''YAYOI'' and the one shot TRIGA ''NSRR'' with the measuring systems using those principles. (auth.)

  19. Neutron-irradiation facilities at the Intense Pulsed Neutron Source-I for fusion magnet materials studies

    International Nuclear Information System (INIS)

    Brown, B.S.; Blewitt, T.H.

    1982-01-01

    The decommissioning of reactor-based neutron sources in the USA has led to the development of a new generation of neutron sources that employ high-energy accelerators. Among the accelerator-based neutron sources presently in operation, the highest-flux source is the Intense Pulsed Neutron Source (IPNS), a user facility at Argonne National Laboratory. Neutrons in this source are produced by the interaction of 400 to 500 MeV protons with either of two 238 U target systems. In the Radiation Effects Facility (REF), the 238 U target is surrounded by Pb for neutron generatjion and reflection. The REF has three separate irradiation thimbles. Two thimbles provide irradiation temperatures between that of liquid He and several hundred degrees centigrade. The third thimble operates at ambient temperature. The large irradiation volume, the neutron spectrum and flux, the ability to transfer samples without warm up, and the dedication of the facilities during the irradiation make this ideally suited for radiation damage studies on components for superconducting fusion magnets. Possible experiments for fusion magnet materials are discussed on cyclic irradiation and annealing of stabilizers in a high magnetic field, mechanical tests on organic insulation irradiated at 4 K, and superconductors measured in high fields after irradiation

  20. Mathematical modelling and quality indices optimization of automatic control systems of reactor facility

    International Nuclear Information System (INIS)

    Severin, V.P.

    2007-01-01

    The mathematical modeling of automatic control systems of reactor facility WWER-1000 with various regulator types is considered. The linear and nonlinear models of neutron power control systems of nuclear reactor WWER-1000 with various group numbers of delayed neutrons are designed. The results of optimization of direct quality indexes of neutron power control systems of nuclear reactor WWER-1000 are designed. The identification and optimization of level control systems with various regulator types of steam generator are executed

  1. Berkeley Nuclear Laboratories Reactor Physics Mk. III Experimental Programme. Description of facility and programme for 1971

    Energy Technology Data Exchange (ETDEWEB)

    Nunn, R M; Waterson, R H; Young, J D

    1971-01-15

    Reactor physics experiments have been carried out at Berkeley Nuclear Laboratories during the past few years in support of the Civil Advanced Gas-Cooled Reactors (Mk. II) the Generating Board is building. These experiments are part of an overall programme whose objective is to assess the accuracy of the calculational methods used in the design and operation of these reactors. This report provides a description of the facility for the Mk. III experimental programme and the planned programme for 1971.

  2. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  3. Operating manual for the High Flux Isotope Reactor. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    None

    1965-06-01

    This report contains a comprehensive description of the High Flux Isotope Reactor facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procedures are presented in another report.

  4. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    International Nuclear Information System (INIS)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

  5. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  6. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    International Nuclear Information System (INIS)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

  7. Computer simulation of radiation processes in reactor facilities

    International Nuclear Information System (INIS)

    Gann, V.V.; Abdulaev, A.M.; Zhukov, A.I.; Marekhin, S.V.; Soldatov, S.A.

    2009-01-01

    The paper describes experience of the code system ALPHA-H/PHOENIX-H/ANC-H (APA) and the code MCNP usage for fuel assembly neutronic calculations and modeling of VVER-1000 reactor core. Using Monte Carlo code MCNP, calculations of neutron field and pin-by-pin energy deposition distributions are provided for different type of assemblies in reactor core. An MCNP model for unit 3 Zaporozhye NPP reactor core was designed. Calculations for pin-by-pin energy deposition in the reactor core were performed using the code system APA and the code MCNP. Comparison of these calculations shows rather high precision of APA calculation for energy deposition in the fuel rods and assemblies operated in the reactor core

  8. Education and research at the VR-1 Vrabec training reactor facility

    International Nuclear Information System (INIS)

    Matejka, K.

    1993-01-01

    The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

  9. A new digital pulse power supply in heavy ion research facility in Lanzhou

    Science.gov (United States)

    Wang, Rongkun; Chen, Youxin; Huang, Yuzhen; Gao, Daqing; Zhou, Zhongzu; Yan, Huaihai; Zhao, Jiang; Shi, Chunfeng; Wu, Fengjun; Yan, Hongbin; Xia, Jiawen; Yuan, Youjin

    2013-11-01

    To meet the increasing requirements of the Heavy Ion Research Facility in Lanzhou-Cooler Storage Ring (HIRFL-CSR), a new digital pulse power supply, which employs multi-level converter, was designed. This power supply was applied with a multi H-bridge converters series-parallel connection topology. A new control model named digital power supply regulator system (DPSRS) was proposed, and a pulse power supply prototype based on DPSRS has been built and tested. The experimental results indicate that tracking error and ripple current meet the requirements of this design. The achievement of prototype provides a perfect model for HIRFL-CSR power supply system.

  10. Responses of the biogas process to pulses of oleate in reactors treating mixtures of cattle and pig manure

    DEFF Research Database (Denmark)

    Nielsen, Henrik Bjørn; Ahring, Birgitte Kiær

    2006-01-01

    The effect of oleate on the anaerobic digestion process was investigated. Two thermophilic continuously stirred tank reactors (CSTR) were fed with mixtures of cattle and pig manure with different total solid (TS) and volatile solid (VS) content. The reactors were subjected to increasing pulses...

  11. Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited)

    International Nuclear Information System (INIS)

    Smith, Roger J.

    2008-01-01

    A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B pol diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T e , n e , and B || along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n e B || product and higher n e and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

  12. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion

    International Nuclear Information System (INIS)

    Marques, J.G.; Sousa, M.; Santos, J.P.; Fernandes, A.C.

    2011-01-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1 MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  13. Environmental assessment for the deactivation of the N Reactor facilities. Revision 1

    International Nuclear Information System (INIS)

    1994-11-01

    This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. The N Reactor facilities are currently in a surveillance and maintenance program, and will eventually be decontaminated and decommissioned (D and D). Operation and maintenance of the facilities resulted in conditions that could adversely impact human health or the environment if left as is until final D and D. The Proposed Action would deactivate the facilities to remove the conditions that present a potential threat to human health and the environment and to reduce surveillance and maintenance requirements. The action would include surveillance and maintenance after deactivation. Deactivation would take about three years and would involve about 80 facilities. Surveillance and maintenance would continue until final D and D, which is expected to be complete for all facilities except the N Reactor itself by the year 2018

  14. An analysis of decommissioning costs for the AFRRI TRIGA reactor facility

    International Nuclear Information System (INIS)

    Forsbacka, Matt

    1990-01-01

    A decommissioning cost analysis for the AFRRI TRIGA Reactor Facility was made. AFRRI is not at this time suggesting that the AFRRI TRIGA Reactor Facility be decommissioned. This report was prepared to be in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations which requires the assurance of availability of future decommissioning funding. The planned method of decommissioning is the immediate decontamination of the AFRRI TRIGA Reactor site to allow for restoration of the site to full public access - this is called DECON. The cost of DECON for the AFRRI TRIGA Reactor Facility in 1990 dollars is estimated to be $3,200,000. The anticipated ancillary costs of facility site demobilization and spent fuel shipment is an additional $600,000. Thus the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for this cost estimate is a study of the decommissioning costs of a similar reactor facility that was performed by Battelle Pacific Northwest Laboratory (PNL) as provided in USNRC publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA. (author)

  15. Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

    2000-12-01

    Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

  16. A description of the Canadian irradiation-research facility proposed to replace the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, A G; Lidstone, R F; Bishop, W E; Talbot, E F; McIlwain, H [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-31

    To replace the aging NRU reactor, AECL has developed the concept for a dual-purpose national Irradiation Research Facility (IRF) that tests fuel and materials for CANDU (CANada Deuterium Uranium) reactors and performs materials research using extracted neutron beams. The IRF includes a MAPLE reactor in a containment building, experimental facilities, and support facilities. At a nominal reactor power of 40 MW{sub t}, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 10{sup 18} n{center_dot}m{sup -2}{center_dot}s{sup -1} in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs.

  17. Engineering evaluation/cost analysis for the 105-DR and 105-F Reactor facilities and ancillary facilities

    International Nuclear Information System (INIS)

    Coenenberg, E.T.

    1998-01-01

    This document presents the results of an engineering evaluation/cost analysis (EE/CA) that was conducted to evaluate alternatives to address final disposition of the 105-DR and 105-F Reactor Buildings (subsequently referred to as facilities), including the fuel storage basins (FSB) and below-grade portions of the reactors, excluding the reactor blocks. The reactor blocks will remain in a safe storage mode for up to 75 years as identified in the Record the Decision (ROD) (58 FR 48509) for the Environmental Impact Statement (EIS), Decommissioning of Eight Surplus Production Reactors at the Hanford Site, Richland, Washington (DOE 1992a). This EE/CA also addresses final disposition of four ancillary facilities: 116-D and 116-DR Exhaust Air Stacks, 117-DR Exhaust Filter Building, and 119-DR Exhaust Air Sample Building. The 105-DR and 105-F facilities are located in the 100-D and 100-F Areas of the Hanford Site. In November 1989, the 100 Area of the Hanford Site was placed on the U.S. Environmental Protection Agency's (EPA) National Priorities List (NPL) under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). The 100 Area NPL includes the 100-D Area (which includes the 100-DR site) and the 100-F Area, which are in various stages of the remediation process. It has been determined by DOE that hazardous substances in the 105-DR, 105-F, and the four ancillary facilities may present a potential threat to human health or the environment, and that a non-time critical removal action at these facilities is warranted. To help determine the most appropriate action, DOE, in cooperation with the Washington State Department of Ecology (Ecology) and the EPA, has prepared this EE/CA. The scope of the evaluation includes the 105-DR and 105-F facilities and the four ancillary facilities. The 116-DR and 117-DR facilities are located within the boundaries of the 105-DR Large Sodium Fire Facility Treatment, Storage, and Disposal (TSD) unit, which is

  18. Situation of the radioactive waste management and the employee radiation exposure in commercial power generation reactor facilities in fiscal 1980

    International Nuclear Information System (INIS)

    1981-01-01

    (1) Situation of the radioactive waste management in commercial power generating reactor facilities: The owners of power generation reactor facilities are obligated not to exceed the target dose around the sites by law in the radioactive waste management. The release of radioactive gaseous and liquid wastes and the storage of radioactive solid wastes in respective reactor facilities in fiscal 1980 are presented in tables (for the former, the data since 1971 are also given). The release control values were satisfied in all the facilities. (2) Situation of employe radiation exposure in commercial power generating reactor facilities: The owners of power generation reactor facilities are obligated not to exceed the permissible exposure doses by law. The Employe exposure doses in respective reactor facilities in fiscal 1980 are given in tables. All exposure doses were below the permissible levels. (J.P.N.)

  19. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    International Nuclear Information System (INIS)

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed

  20. A Design of Alarm System in a Research Reactor Facility

    International Nuclear Information System (INIS)

    Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk

    2013-01-01

    The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants

  1. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.; Ott, K.O.; Terry, W.K.

    1987-01-01

    A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

  2. Progress on ANSTO'S OPAL reactor project and its future importance as the centrepiece of ANSTO'S facilities

    International Nuclear Information System (INIS)

    Smith, I.O.

    2006-01-01

    Full text: After an intensive process of analysis, the Australian government approved the construction of a multi-purpose research reactor in 1997. Following the conduct of a comprehensive tender evaluation process in 1998-2000, INVAP was contracted to construct a 20 MW open pool research reactor and associated neutron beam facilities. The construction of the reactor is now almost complete, and we have commenced cold commissioning. ANSTO has applied for an operating licence, and we hope for a decision on that application in June, following the consideration by the regulator of the results of cold commissioning. The OPAL reactor will provide neutrons to a world-class neutron beam facility, in which a number of the instruments will have the best performance available in the world to date. We intend to establish the Bragg Institute as a regional centre of excellence on neutron beam science, with a significant number of international scientists using the facility to produce cutting edge science in the fields of biology, materials science, food science and other area. The reactor also has extensive irradiation facilities within the reflector vessel. These facilities will be used to produce medical isotopes - ANSTO supplies the bulk of the Australian market and also exports into this region - and for the transmutation doping of silicon ingots for semiconductor manufacture. There are also a number of pneumatically loaded radiation facilities allowing for short term irradiation of samples for such activities as neutron activation analysis

  3. Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

    Directory of Open Access Journals (Sweden)

    Sayer C.

    2002-01-01

    Full Text Available A new reactor, the pulsed sieve plate column (PSPC, was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA emulsion polymerization reactions performed in this PSPC. Therefore, both experimental studies and reaction simulations were performed. Results showed that it is possible to obtain high conversions with rather low residence times in the PSPC.

  4. Design of a 28 MW pulse facility for testing superconducting coils to several hundred megajoules capacity

    International Nuclear Information System (INIS)

    Vogel, H.F.

    1980-01-01

    Railway traction motors are available in unit sizes convenient for installation and series-parallel grouping. They are rugged. Industry builds and refurbishes them with good economy and in quantities replenishing the rolling stock. We find them well suited for reversing the current in a superconducting winding. We focus on a pulsed energy of 20 to 100 MJ, discussing our analysis and facility planning. Limitations are imposed by the following maximum numbers tolerated by the motor - pulsed current of 3.0 to 3.5 kA, current change of 40 kA/s, and pulsed voltage of 1.8 kV. Hence, the number of machines needed in parallel follows from the coil current and its rate of change. The number in series is determined by the voltage. The power transfer is limited by the torsional strength of the motor shaft to a value affected by the flywheel mass

  5. Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han

    2013-01-01

    The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility

  6. Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility.

  7. The upgrading of the cyclic neutron activation analysis facility at the Dalat research reactor

    International Nuclear Information System (INIS)

    Van Doanh Ho; Manh Dung Ho; Quang Thien Tran; Dong Vu Cao; Thanh Viet Ha

    2018-01-01

    The cyclic neutron activation analysis (CNAA) facility based on a pneumatic transfer system for short irradiation and rapid counting has recently been upgraded at the Dalat research reactor. The original facility was only designed for single irradiation. Therefore, this work has aimed to upgrade both hardware and software for the cyclic irradiation. In this paper, the upgrading of the facility for CNAA was described. Irradiation time of the facility were calibrated, thereby reducing irradiation time to seconds with precision. The accuracy and sensitivity of CNAA based-on the upgraded facility were assessed by determination of some short-lived nuclides. (author)

  8. Public's right to information: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Stokely, E.

    1981-02-01

    The events at TMI prompted the Under Secretary of the Department of Energy (DOE) to establish the Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee. This Committee was assigned the task of assessing the adequacy of nuclear facility personnel qualification and training at DOE-owned reactors in light of the Three Mile Island accident. The Committee was also asked to review recommendations and identify possible implications for DOE's nuclear facilities

  9. NRX and NRU reactor research facilities and irradiation and examination charges

    International Nuclear Information System (INIS)

    1960-08-01

    This report details the irradiation and examination charges on the NRX and NRU reactors at the Chalk River Nuclear Labs. It describes the NRX and NRU research facilities available to external users. It describes the various experimental holes and loops available for research. It also outlines the method used to calculate the facilities charges and the procedure for applying to use the facilities as well as the billing procedures.

  10. A pulsed supersonic gas jet target for precision spectroscopy at the HITRAP facility at GSI

    Science.gov (United States)

    Tiedemann, D.; Stiebing, K. E.; Winters, D. F. A.; Quint, W.; Varentsov, V.; Warczak, A.; Malarz, A.; Stöhlker, Th.

    2014-11-01

    A pulsed supersonic gas jet target for experiments at the HITRAP facility at the GSI Helmholtzzentrum für Schwerionenforschung in Darmstadt has been designed and built as a multi-purpose installation for key experiments on fundamental atomic physics in strong fields. This setup is currently installed at the Institut für Kernphysik of Goethe-University, Frankfurt am Main (IKF), in order to explore its operation prior to its installation at the HITRAP facility. Design and performance of the target are described. The measured target densities of 5.9×1012 atoms/cm3 for helium and 8.1×1012 atoms/cm³ for argon at the stagnation pressure of 30 bar match the required values. The target-beam diameter of 0.9 mm and the pulsed operation mode (jet built-up-time ≤15 ms) are well suited for the use at HITRAP.

  11. Advanced Test Reactor National Scientific User Facility 2010 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Mary Catherine Thelen; Todd R. Allen

    2011-05-01

    This is the 2010 ATR National Scientific User Facility Annual Report. This report provides an overview of the program for 2010, along with individual project reports from each of the university principal investigators. The report also describes the capabilities offered to university researchers here at INL and at the ATR NSUF partner facilities.

  12. Implementation of safeguards and security for fissile materials disposition reactor alternative facilities

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

    1995-01-01

    A number of different disposition alternatives are being considered and include facilities which provide for long-ten-n and interim storage, convert and stabilize fissile materials for other disposition alternatives, immobilize fissile material in glass and/or ceramic material, fabricate fissile material into mixed oxide (MOX) fuel for reactors, use reactor based technologies to convert material into spent fuel, and dispose of fissile material using a number of geologic alternatives. Particular attention will be given to the reactor alternatives which include existing, partially completed, advanced or evolutionary LWRs and CANDU reactors. The various reactor alternatives are all very similar and include processing which converts Pu to a usable form for fuel fabrication, a MOX fuel fab facility located in either the US or in Europe, US LWRs or the CANDU reactors and ultimate disposal of spent fuel in a geologic repository. This paper focuses on how the objectives of reducing security risks and strengthening arms reduction and nonproliferation will be accomplished and the possible impacts of meeting these objectives on facility operations and design. Some of the areas in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threat, and (4) potential proliferation risks, the impacts on the facilities, and safeguards and security issues unique to the presence of Category 1 or strategic special nuclear material

  13. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    International Nuclear Information System (INIS)

    Bissani, M; O'Kelly, D S

    2006-01-01

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

  14. Combined treatment of SO2 and high resistivity fly ash using a pulse energized electron reactor

    International Nuclear Information System (INIS)

    Mizuno, A.; Clements, J.S.; Davis, R.H.

    1984-01-01

    The combined removal of SO 2 and high resistivity fly ash has been demonstrated in a pulse energized electron reactor (PEER). The PEER system which was originally developed for the removal of SO 2 utilizes a positive pulse streamer corona discharge in a non-uniform field geometry. In performance tests on SO 2 , more than 90% was removed with an advantageously small power requirement. Combined treatment performance was demonstrated by introducing high resistivity fly ash into the test gas and the PEER is significantly more efficient than a conventional electrostatic precipitator operated with a dc voltage. Observations show that the PEER agglomerates the fly ash and further that the SO 2 removal efficiency is improved by the presence of fly ash. The electrode configuration and performance results make retrofit consideration attractive

  15. Study of two-zone reactor system using a pulsed neutron technique

    International Nuclear Information System (INIS)

    Shishin, B.P.; Platovskikh, Yu.A.; Didejkin, T.S.

    1977-01-01

    Theoretical and experimental investigations of a neutron flux time dependence after a sport fast neutron pulse in a reactor core - neutron reflector multiplying system have been conducted. A correlation between eigenvalues governing neutron flux decrease at t→infinity for the two-zone system and eigenvalues for each zone has been established in terms of the one-group diffusion approximation. Experiments have been performed in an experimental subcritical assembly comprising a cylindrical uranium core surrounded by a radial water reflector with different boric acid concentrations. The experiments show that the observed neutron flux decrease in the core is governed by an exponent exp(-Λ 1 t), whereas in the reflector by a sum of two exponents exp(-Λ 1 t) and exp(-Λ 2 t). The eigenvalue Λ 1 reflects multiplying properties of the reactor, and Λ 2 is determined by the reflector absorption cross section

  16. The analysis of neutron physical characteristics of fast reactors by means of pulsed experiments

    International Nuclear Information System (INIS)

    Stumbur, Eh.A.; Milyutina, Z.N.

    1992-01-01

    Possibility is considered for determination of macroscopic cross sections of homogeneous multiplicating media with fast neutrons. It is shown that by means of the critical size, laplaccian and neutron pulse damping decrement measurement results it is possible to obtain values of almost all cross sections of a medium. The method is tested with systems of metal 235 U and BFS-32 assemblies with the composition, typical for fast power reactors. A suitable algorithm is developed for solving nonstationary asymptotic transport problems. Calculation results are compared with experimental ones. 21 refs.; 2 figs.; 3 tabs

  17. A low background pulsed neutron polyenergetic beam at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Adib, M.; Abdel-Kawy, A.; Habib, N.; Abu-El-Ela, M.; Wahba, M.; Kilany, M.

    1991-12-01

    A low background pulsed neutron polyenergetic thermal beam at ET-RR-1 is produced by a rotor and rotating collimator suspended in magnetic fields. Each of them is mounted on its mobile platform and whose centres are 66 cm apart, rotating synchronously at speeds up to 16000 rpm. It was found that the neutron burst produced by the rotor with almost 100% transmission passes through the collimator, when the rotation phase between them is 28.8 deg. Moreover the background level achieved at the detector position is low, constant and free from peaks due to gamma rays and fast neutrons accompanying the reactor thermal beam. (author). 12 refs, 3 figs

  18. Study of two-zone reactor system using a pulsed neutron technique

    Energy Technology Data Exchange (ETDEWEB)

    Shishin, B P; Platovskikh, Yu A; Didejkin, T S

    1977-05-01

    Theoretical and experimental investigations of a neutron flux time dependence after a sport fast neutron pulse in a reactor core - neutron reflector multiplying system have been conducted. A correlation between eigenvalues governing neutron flux decrease at t..-->..infinity for the two-zone system and eigenvalues for each zone has been established in terms of the one-group diffusion approximation. Experiments have been performed in an experimental subcritical assembly comprising a cylindrical uranium core surrounded by a radial water reflector with different boric acid concentrations.

  19. Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

    Energy Technology Data Exchange (ETDEWEB)

    McKenzie, J L [Armed Forces Radiobiology Research Institute, Bethesda, MD (United States)

    1974-07-01

    The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the {sup 41}Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

  20. Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

    International Nuclear Information System (INIS)

    McKenzie, J.L.

    1974-01-01

    The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the 41 Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

  1. Study of startup conditions of a pulsed annular reactor; Estudo das reacoes de partida de um reator anelar pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mario Augusto Bezerra da

    2003-10-15

    A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

  2. Fast reactor test facilities in the US safety program

    International Nuclear Information System (INIS)

    Avery, R.; Dickerman, C.E.; Lennox, D.H.; Rose, D.

    1979-01-01

    The needs for safety information derivable from in-pile programs are reviewed, and the correlation made with existing and planned capability. In view of the current status of the U.S. breeder program, emphasis is given in the review to the impact of different fast breeder options on the required program and facilities. It is concluded that facility needs are somewhat independent of specific fast breeder concept, even though the relative emphasis on the various safety issues will differ. 8 refs

  3. Experimental measurements and theoretical simulations for neutron flux in self-serve facility of Dhruva reactor

    International Nuclear Information System (INIS)

    Rana, Y.S.; Mishra, Abhishek; Singh, Tej

    2016-06-01

    Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)

  4. Emergency planning and response: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Knuth, D.; Boyd, R.

    1981-02-01

    The Department of Energy (DOE) has formed a Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee to assess the implications of the recommendations contained in the President's Commission Report on the Three Mile Island (TMI) Accident (the Kemeny Commission report) that are applicable to DOE's nuclear reactor operations. Thirteen DOE nuclear reactors have been reviewed. The assessments of the 13 facilities are based on information provided by the individual operator organizations and/or cognizant DOE Field Offices. Additional clarifying information was supplied in some, but not all, instances. This report indicates how these 13 reactor facilities measure up in light of the Kemeny and other TMI-related studies and recommendations, particularly those that have resulted in upgraded Nuclear Regulatory Commission (NRC) requirements in the area of emergency planning and response

  5. Cold neutron PGAA facility developments at university research reactors in the USA

    International Nuclear Information System (INIS)

    Uenlue, K.; Rios-Martinez, C.

    2005-01-01

    The PGAA applications can be enhanced by using subthermal neutrons, cold neutrons at university research reactors. Only two cold neutron beam facilities were developed at the U.S. university research reactors, namely at Cornell University and the University of Texas at Austin. Both facilities used mesitylene moderator. The mesitylene moderator in the Cornell Cold Neutron Beam Facility (CNBF) was cooled by a helium cryorefrigerator via copper cold fingers to maintain the moderator below 30 K at full power reactor operation. Texas Cold Neutron Source (TCNS) also uses mesitylene moderator that is cooled by a cryorefrigerator via a neon thermosiphon. The operation of the TCNS is based on a helium cryorefrigerator, which liquefies neon gas in a 3-m long thermosiphon. The thermosiphon cools and maintains mesitylene moderator at about 30 K in a chamber. Neutrons streaming through the mesitylene chamber are moderated and thus reduce their energy to produce a cold neutron distribution. (author)

  6. Self-sustainability of a research reactor facility with neutron activation analysis

    International Nuclear Information System (INIS)

    Chilian, C.; Kennedy, G.

    2010-01-01

    Long-term self-sustainability of a small reactor facility is possible because there is a large demand for non-destructive chemical analysis of bulk materials that can only be achieved with neutron activation analysis (NAA). The Ecole Polytechnique Montreal SLOWPOKE Reactor Facility has achieved self-sustainability for over twenty years, benefiting from the extreme reliability, ease of use and stable neutron flux of the SLOWPOKE reactor. The industrial clientele developed slowly over the years, mainly because of research users of the facility. A reliable NAA service with flexibility, high accuracy and fast turn-around time was achieved by developing an efficient NAA system, using a combination of the relative and k0 standardisation methods. The techniques were optimized to meet the specific needs of the client, such as low detection limit or high accuracy at high concentration. New marketing strategies are presented, which aim at a more rapid expansion. (author)

  7. Accidents and failures in reactor facilities for test and research and reactor facilities in the stage of research and development in fiscal year 1987

    International Nuclear Information System (INIS)

    1988-01-01

    The number of accidents and failures reported in fiscal year 1987 in conformity with the law on the regulation of nuclear reactors and others was three. One case occurred during operation, and two cases occurred in shutdown state. One case was caused by improper construction management, and two cases were due to improper maintenance management. The effect of radioactivity to the surrounding environment of reactor facilities due to these accidents and failures did not arise. These occurred in the NSRR of Japan Atomic Energy Research Institute (Tokai), the experimental FBR Joyo and the ATR Fugen Power Station of Power Reactor and Nuclear Fuel Development Corp. In addition to these, the light troubles reported on the basis of the notice from the director of Science and Technology Agency dated September 1, 1981, were three cases. (K.I.)

  8. Sharing of the RPI Reactor Critical Facility (RCF). Final summary report, January 1988--September 1995

    International Nuclear Information System (INIS)

    Harris, D.R.

    1995-01-01

    Rensselaer Polytechnic Institute (RPI) has participated for a number of years in Sharing of the Reactor Critical Facility (RCF) under the U.S. Department of Energy University Reactor Sharing Program. In September of each year a Sharing invitation is sent to 92 public and private high schools and to 74 colleges and universities within about a 3 hour drive to the RCF (Appendix B). Each year about 10 such educational institutions send groups to share the RCF

  9. Experimental facility of innovative types as the laboratory analog of research reactor experimental device

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Zabud'ko, A.N.; Kremenetskij, A.K.; Nikolaev, A.N.; Trykov, L.A.

    1991-01-01

    The paper analyses capability of creating laboratory analogs of complex experimental facilities at research reactors utilizing power radionuclide neutron sources fabricated in industrial conditions. Some experimental and calculational investigations of neutron-physical characteristics are presented, which have been attained at the RIZ research reactor laboratory analog. Experimental results are supplemented by calculational investigations, fulfilled by means of the BRAND three-dimensional computational complex and the ROZ-6 one-dimensional program. 4 refs.; 3 figs

  10. The regulation and licensing of research reactors and associated facilities in the United Kingdom

    International Nuclear Information System (INIS)

    Weightman, M.W.; Willby, C.R.

    1990-01-01

    In the United Kingdom, the Nuclear Installations Inspectorate (NII) licenses nuclear facilities, including research reactors, on behalf of the Health and Safety Executive (HSE). The legislation, the regulatory organizations and the methods of operation that have been developed over the last 30 years result in a largely non-prescriptive form of control that is well suited to research reactors. The most important part of the regulatory system is the license and the attachment of conditions which it permits. These conditions require the licensee to prepare arrangements to control the safety of the facility. In doing so the licensee is encouraged to develop a 'safety culture' within its organization. This is particularly important for research reactors which may have limited staff resources and where the ability, and at times the need, to have access to the core is much greater than for nuclear power plants. Present day issues such as the ageing of nuclear facilities, public access to the rationale behind regulatory decisions, and the emergence of more stringent safety requirements, which include a need for quantified safety criteria, have been addressed by the NII. This paper explores the relevance of such issues to the regulation of research reactors. In particular, it discusses some of the factors associated with research reactors that should be considered in developing criteria for the tolerability of risk from these nuclear facilities. From a consideration of these factors, it is the authors' view that the range of tolerable risk to the public from the operation of new research reactors may be expected to be more stringent than similar criteria for new nuclear power plants, whereas the criteria for tolerable risk for research reactor workers are expected to be about the same as those for power reactor workers

  11. Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    2002-01-01

    A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

  12. Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited).

    Science.gov (United States)

    Smith, Roger J

    2008-10-01

    A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B(pol) diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T(e), n(e), and B(parallel) along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n(e)B(parallel) product and higher n(e) and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

  13. Optimization of a neutron detector system for the fourier RTOF-diffractometer facility at the ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maayouf, R M.A.; El-Kady, A S.I.; El-Shaer, Y H [Reactor and Neutron physics Dept., NRC, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    This work deals with optimization of a detector system for use with the fourier reverse time of flight (RTOF) diffractometer facility recently installed at the ETRR-1 reactor. The detector system of the diffractometer was designed for detecting neutrons scattered from the sample. It consists of 4 independent scintillation detector elements arranged according to the time focusing geometry in order to increase the luminosity of the diffractometer for the given resolution. Each of the detector elements is made of 1 mm thick {sup 6} Li-glass scintillator (NE-912) whose surface area is 200 mm{sup 2}. The present detector system ensures reliable identification and separation of thermal neutron pulses from fast neutrons and gamma- quanta. It was found from measurements with Fe sample that it is possible to make neutron diffraction measurement, within one hour using the present detector system and with resultion 0.5%. 6 FIGS.

  14. Detailed description of an SSAC at the facility level for on-load refueled power reactor facilities

    International Nuclear Information System (INIS)

    Jones, R.J.

    1985-11-01

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in an on-load refueled power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  15. Heat removing device for nuclear reactor container facility

    Energy Technology Data Exchange (ETDEWEB)

    Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

    1994-09-30

    A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

  16. Pulse*Star Inertial Confinement Fusion Reactor: heat transfer loop and balance of plant considerations

    International Nuclear Information System (INIS)

    McDowell, M.W.; Murray, K.A.

    1984-01-01

    A conceptual heat transfer loop and balance of plant design for the Pulse*Star Inertial Confinement Fusion Reactor has been investigated and results are presented. The Pulse*Star reaction vessel, a perforated steel bell jar approximately 11 m in diameter, is immersed in Li 17 Pb 83 coolant which flows through the perforations and forms a 1.5 m thick plenum of droplets around an 8 m diameter inner chamber. The reactor and associated pumps, piping, and steam generators are contained within a 17 m diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops with flow rates of 5.5 m 3 /s each are necessary to transfer 3300 MWt of power. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. Power balance calculations based on an improved electrical conversion efficiency revealed a net electrical output of 1260 MWe to the bus bar and a resulting net efficiency of 39%. Suggested balance-of-plant layouts are also presented

  17. Performance Variation of Spent Resin in Mixed Bed From Water Purifying System of Xi'an Pulse Reactor

    International Nuclear Information System (INIS)

    Li Hua; Ma Yan; Xiao Yan; Liu Yueheng; Yang Yongqing

    2010-01-01

    Detailed physical and chemical characteristic analysis was performed on the spent cation and anion resins in the mixed bed from Xi'an Pulse Reactor water purifying system.The exchange performance variations of used resins and the contributions from different factors to the variation were discussed.Based on the obtained information of the impurities in the used resin, the contamination state of the water in the Xi'an Pulse Reactor water pool, the corrosion state of the structural material in the reactor was presented. The spent anion resin almost completely losses its exchange performance,while the remaining exchange capacity in the spent cation resin is still high.The radiation field from the reactor operation contributes little to the degradation of the performance of the resins. The exchange capacity loss of the spent anion resin is due to the exchange of its active groups into abundant carbonate and a certain amount of organics. The impurity amount in the anion and cation exchange resins is low,which suggests(that) the water in the Xi'an Pulse Reactor water pool is little contaminated. A certain extent of corrosion is occurred on the structural material in the swimming pool of the reactor. The results provide important referential data for the operational safety of the water purifying system of similar research reactor. (authors)

  18. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    International Nuclear Information System (INIS)

    Salazar, M.D.

    1998-01-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel

  19. Development programs on decommissioning technology for reactors and fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Fujiki, K.

    1992-01-01

    The Science and Technology Agency (STA) of Japan is promoting technology development for decommissioning of nuclear facilities by entrusting various research programs to concerned research organisations: JAERI, PNC and RANDEC, including first full scale reactor decommissioning of JPDR. According to the results of these programs, significant improvement on dismantling techniques, decontamination, measurement etc. has been achieved. Further development of advanced decommissioning technology has been started in order to achieve reduction of duration of decommissioning work and occupational exposures in consideration of the decommissioning of reactors and fuel cycle facilities. (author) 5 refs.; 7 figs.; 1 tab

  20. Physics design of fast reactor safety test facilities for in-pile experiments

    International Nuclear Information System (INIS)

    Travelli, A.; Matos, J.E.; Snelgrove, J.L.; Shaftman, D.H.; Tzanos, C.P.; Lam, S.K.; Pennington, E.M.; Woodruff, W.L.

    1976-01-01

    A determined effort to identify and resolve current Fast Breeder Reactor safety testing needs has recently resulted in a number of conceptual designs for FBR safety test facilities which are very complex and diverse both in their features and in their purpose. The paper discusses the physics foundations common to most fast reactor safety test facilities and the constraints which they impose on the design. The logical evolution, features, and capabilities of several major conceptual designs are discussed on the basis of this common background

  1. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    Energy Technology Data Exchange (ETDEWEB)

    Salazar, M.D.

    1998-12-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel.

  2. Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

    International Nuclear Information System (INIS)

    1975-12-01

    The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

  3. Calculation and experimental measurements in the Argonauta reactor subcritical and exponential facility

    International Nuclear Information System (INIS)

    Voi, Dante L.; Furieri, Rosane C.A.A.; Renke, Carlos A.C.; Bastos, Wilma S.; Ferreira, Francisco J.O.

    1997-01-01

    Initial measurements were performed on the exponential and subcritical facility installed on the internal thermal column of the Argonauta reactor at IEN-CNEN-Rio de Janeiro, Brazil. The measurements are include in the reactor physics experimental program for integral parameters determination, for both valid and confirmed theoretical models for reactor calculation. Gamma doses and neutron fluxes were measured with telescopic, proportional counters, wire and foil detectors. Experimental data were compared with results obtained by application of CITATION code. (author). 4 refs., 8 figs

  4. Progress of the ELISE test facility: towards one hour pulses in hydrogen

    Science.gov (United States)

    Wünderlich, D.; Fantz, U.; Heinemann, B.; Kraus, W.; Riedl, R.; Wimmer, C.; the NNBI Team

    2016-10-01

    In order to fulfil the ITER requirements, the negative hydrogen ion source used for NBI has to deliver a high source performance, i.e. a high extracted negative ion current and simultaneously a low co-extracted electron current over a pulse length up to 1 h. Negative ions will be generated by the surface process in a low-temperature low-pressure hydrogen or deuterium plasma. Therefore, a certain amount of caesium has to be deposited on the plasma grid in order to obtain a low surface work function and consequently a high negative ion production yield. This caesium is re-distributed by the influence of the plasma, resulting in temporal instabilities of the extracted negative ion current and the co-extracted electrons over long pulses. This paper describes experiments performed in hydrogen operation at the half-ITER-size NNBI test facility ELISE in order to develop a caesium conditioning technique for more stable long pulses at an ITER relevant filling pressure of 0.3 Pa. A significant improvement of the long pulse stability is achieved. Together with different plasma diagnostics it is demonstrated that this improvement is correlated to the interplay of very small variations of parameters like the electrostatic potential and the particle densities close to the extraction system.

  5. The FRJ 1 reactor (MERLIN) at Juelich, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FRJ 1 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  6. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits

  7. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

  8. Thermal hydraulic modelling of the Mo and Iridium irradiation facilities of the RA10 reactor

    International Nuclear Information System (INIS)

    Gramajo, M.; García, J.; Marcel, C.P.

    2013-01-01

    The RA-10 reactor is a multipurpose, open pool research reactor. The core consists of a rectangular array of MTR type fuel. The produced thermal power is 30 MW which is extracted by the refrigeration system via an ascendant flow through the core. The core reflector is D 2 O contained in a watertight tank. The design of the reactor includes a number of out-core facilities which are meant to be used for industrial, medical and research purposes. Among all the facilities, the most important ones are the Molybdenum and Iridium ones which we modeled in this work. During the normal operation of the reactor, the manipulation and the on-line extraction of the irradiation facilities is foreseen. Therefore the study of the head loss during the normal operation as well as during the extraction maneuvers plays a relevant role in the design and safety analysis. In this work a CFD commercial code is use dto perform the calculations needed to guarantee the design requirements.In addition, a full detailed geometric model for both, the Molybdenum and Iridium facilities,is used to perform the required simulations. The obtained results allow to evaluating the thermal-hydraulic performance of the proposed facilities designs. (author)

  9. Good Practices for Water Quality Management in Research Reactors and Spent Fuel Storage Facilities

    International Nuclear Information System (INIS)

    2011-01-01

    Water is the most common fluid used to remove the heat produced in a research reactor (RR). It is also the most common media used to store spent fuel elements after being removed from the reactor core. Spent fuel is stored either in the at-reactor pool or in away-from-reactor wet facilities, where the fuel elements are maintained until submission to final disposal, or until the decay heat is low enough to allow migration to a dry storage facility. Maintaining high quality water is the most important factor in preventing degradation of aluminium clad fuel elements, and other structural components in water cooled research reactors. Excellent water quality in spent fuel wet storage facilities is essential to achieve optimum storage performance. Experience shows the remarkable success of many research reactors where the water chemistry has been well controlled. In these cases, aluminium clad fuel elements and aluminium pool liners show few, if any, signs of either localized or general corrosion, even after more than 30 years of exposure to research reactor water. In contrast, when water quality was allowed to degrade, the fuel clad and the structural parts of the reactor have been seriously corroded. The driving force to prepare this publication was the recognition that, even though a great deal of information on research reactor water quality is available in the open literature, no comprehensive report addressing the rationale of water quality management in research reactors has been published to date. This report is designed to provide a comprehensive catalogue of good practices for the management of water quality in research reactors. It also presents a brief description of the corrosion process that affects the components of a research reactor. Further, the report provides a basic understanding of water chemistry and its influence on the corrosion process; specifies requirements and operational limits for water purification systems of RRs; describes good practices

  10. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    International Nuclear Information System (INIS)

    Gentile, C.A.; LaMarche, P.H.

    1995-01-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During '' time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems

  11. Electron energy device for LINAC based Pulse Radiolysis Facility of RPCD

    International Nuclear Information System (INIS)

    Toley, M.A.; Shinde, S.J.; Chaudhari, B.B.; Sarkar, S.K.

    2015-07-01

    The pulse radiolysis facility is the experimental centerpiece of the radiation chemistry activities of the Radiation and Photochemistry Division (RPCD) of Bhabha Atomic Research Centre. This facility was created in 1986 which is based on a 7 MeV Linear Electron Accelerator (LINAC) procured from M/s Radiation Dynamics Ltd., UK. The electron energy is one of the principal parameters that influence the dose distribution within the sample irradiated with a beam of energetic electrons. An easy-to-use and robust device has been developed that can reliably detect day-today small variations in the beam energy. It consists of two identical aluminum plates except for their thickness, which are electrically insulated from each other. The thickness of each plate is carefully selected depending on the electron beam energy. The charge (or current) collected by each plate, under irradiation is measured. The ratio of the charge (or current) signal from the front plate to the sum of the signals from the front and rear plates is very sensitive to the beam energy. The high sensitivity and robustness make this device quite suitable for Electron energy measurement for Pulse radiolysis Facility at RPCD. (author)

  12. 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

  13. A novel pulsed corona discharge reactor based on surface streamers for diesel exhaust remediation

    Energy Technology Data Exchange (ETDEWEB)

    Malik, M.A.; Schoenbach, K.H. [Old Dominion Univ., Norfolk, VA (United States). Frank Reidy Research Center for Bioelectrics

    2010-07-01

    Modelling of surface streamers along insulating surfaces to determine the dielectric strength of insulators in high voltage systems has shown that surface streamers consist of a positive streamer head followed by quasi-neutral plasma in the channel behind and surrounded by a layer of positive charges. This paper described a novel pulsed corona discharge reactor which utilized such surface streamers along insulating surfaces. The electrodes were comprised of a stainless steel wire anode of 150 mm in diameter stretched along the surface of a glass sheet and two parallel aluminum strips as cathodes. An eight-stage Marx bank, was used to produce the surface streamers in nitrogen-oxygen mixtures at atmospheric pressure. The paper described the experimental study with particular reference to the schematics of a surface streamer plasma reactor and the dimensions of discharge spaces of three reactors. The purpose of the study was to find the optimum conditions for energy yield and effective destruction of nitrogen oxides from diesel engine exhaust. It was concluded that surface streamers generate a more diffuse plasma. Energy costs for production of ozone or nitrogen dioxide that require reactions with bulk gas molecules were nearly the same in surface streamer discharges as in volume streamer discharges. 12 refs., 1 tab., 7 figs.

  14. Upgrading of neutron radiography/tomography facility at research reactor

    International Nuclear Information System (INIS)

    Abd El Bar, Waleed; Mongy, Tarek

    2014-01-01

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  15. Seismic qualification of safety class components in non-reactor nuclear facilities at Hanford site

    International Nuclear Information System (INIS)

    Ocoma, E.C.

    1989-01-01

    This paper presents the methods used during the walkdowns to compile as-built structural information to seismically qualify or verify the seismic adequacy of safety class components in the Plutonium Finishing Plant complex. The Plutonium finishing Plant is a non-reactor nuclear facility built during the 1950's and was designed to the Uniform Building Code criteria for both seismic and wind events. This facility is located at the US Department of Energy Hanford Site near Richland, Washington

  16. Status and future program of reactor physics experiments in JAERI Critical facilities, FCA and TCA

    International Nuclear Information System (INIS)

    Okajima, Shigeaki; Osugi, Toshitaka; Nakajima, Ken; Suzaki, Takenori; Miyoshi, Yoshinori

    1999-01-01

    The critical facilities in JAERI, FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly), have been used to provide integral data for evaluation of nuclear data as well as for development of various types of reactor since they went critical in 1960's. In this paper a review is presented on the experimental programs in both facilities. And the experimental programs in next 5 years are also shown. (author)

  17. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    International Nuclear Information System (INIS)

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition

  18. Aqueous Boric acid injection facility of PWR type reactor

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi; Iwami, Masao.

    1996-01-01

    If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

  19. 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.)

  20. Seed irradiation facilities at TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Najzer, M.

    1972-01-01

    Fast neutrons and gamma-rays with their high and low LET respectively are excellent complementary tools for investigation of the effect of different types of mutations. TRIGA Irradiation Facility and Thermal Column Irradiation Facility were designed and installed for the first time in the TRIGA tank and thermal column respectively. The basic idea of design was the use of depleted uranium as gamma-ray and thermal neutron shield and simultaneously as thermal to fast neutron converter. Low LET radiation, due to direct and thermal neutron capture gamma-rays, is strongly attenuated while fast neutron flux is increased. GIF is made of a cadmium tube inserted in a graphite block. It is located in the central thermal column channel. The basic idea is to convert thermal neutrons to gamma-rays by capture in the cadmium

  1. Education and training at the Rensselaer Polytechnic Institute reactor critical facility

    International Nuclear Information System (INIS)

    Harris, D.R.

    1989-01-01

    The Rensselaer Polytechnic Institute (RPI) Reactor Critical Facility (RCF) has provided hands-on education and training for RPI and other students for almost a quarter of a century. The RCF was built in the 1950s by the American Locomotive Company (ALCO) as a critical facility in which to carry out experiments in support of the Army Package power Reactor (APPR) program. A number of APPRs were built and operated. In the middle 1960s, ALCO went out of business and provided the facility to RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although limited amounts of reactor research, activation analysis, and reactivity assays have been carried out as well. Recently, a U.S. Department of Energy (DOE) upgrade program supported refueling the RCF with 4.81 wt% enriched UO 2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective approach to conversion from high-enrichment bombgrade fuel to low-enrichment fuel. More important, however, is the fact that the new fuel is of current interest for light water power reactors with extended lifetime fuel. Thus, not only are critical reactor experiments being carried out on the fuel but, more importantly, the quality of the education and training has been enhanced

  2. PANDA a multi-purpose thermal-hydraulics facility devoted to nuclear reactor containment safety analysis

    International Nuclear Information System (INIS)

    Paladino, Domenico

    2014-01-01

    This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

  3. Analyses in support of the Laboratory Microfusion Facility and ICF commercial reactor designs

    International Nuclear Information System (INIS)

    Meier, W.R.; Monsler, M.J.

    1988-01-01

    Our work on this contract was divided into two major categories; two thirds of the total effort was in support of the Laboratory Microfusion Facility (LMF), and one third of the effort was in support of Inertial Confinement Fusion (ICF) commercial reactors. This final report includes copies of the formal reports, memoranda, and viewgraph presentations that were completed under this contract

  4. The Text of the Agreement for the Application of Agency Safeguards to United States Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-08-14

    The text of the Agreement between the Agency and the Government of the United States of America for the application of Agency safeguards to United States reactor facilities, which was signed on 15 June 1964 and entered into force on 1 August 1964, is reproduced in this document for the information of all Members.

  5. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24

  6. An automatic device for sample insertion and extraction to/from reactor irradiation facilities

    International Nuclear Information System (INIS)

    Alloni, L.; Venturelli, A.; Meloni, S.

    1990-01-01

    At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

  7. Fissile material detection and control facility with pulsed neutron sources and digital data processing

    International Nuclear Information System (INIS)

    Romodanov, V.L.; Chernikova, D.N.; Afanasiev, V.V.

    2010-01-01

    Full text: In connection with possible nuclear terrorism, there is long-felt need of devices for effective control of radioactive and fissile materials in the key points of crossing the state borders (airports, seaports, etc.), as well as various customs check-points. In International Science and Technology Center Projects No. 596 and No. 2978, a new physical method and digital technology have been developed for the detection of fissile and radioactive materials in models of customs facilities with a graphite moderator, pulsed neutron source and digital processing of responses from scintillation PSD detectors. Detectability of fissile materials, even those shielded with various radiation-absorbing screens, has been shown. The use of digital processing of scintillation signals in this facility is a necessary element, as neutrons and photons are discriminated in the time dependence of fissile materials responses at such loads on the electronic channels that standard types of spectrometers are inapplicable. Digital processing of neutron and photon responses practically resolves the problem of dead time and allows implementing devices, in which various energy groups of neutrons exist for some time after a pulse of source neutrons. Thus, it is possible to detect fissile materials deliberately concealed with shields having a large cross-section of absorption of photons and thermal neutrons. Two models of detection and the control of fissile materials were advanced: 1. the model based on graphite neutrons moderator and PSD scintillators with digital technology of neutrons and photons responses separation; 2. the model based on plastic scintillators and detecting of time coincidences of fission particles by digital technology. Facilities that count time coincidences of neutrons and photons occurring in the fission of fissile materials can use an Am Li source of neutrons, e.g. that is the case with the AWCC system. The disadvantages of the facility are related to the issues

  8. Intensive neutron source based on powerful electron linear accelerator LIA-30 and pulsed nuclear reactor FR-1

    Energy Technology Data Exchange (ETDEWEB)

    Bossamykin, V S; Koshelev, A S; Gerasimov, A I; Gordeev, V S; Grishin, A V; Averchenkov, V Ya; Lazarev, S A; Maslov, G N; Odintsov, Yu M [All-Russian Scientific Research Institute of Experimental Physics, Sarov (Russian Federation)

    1997-12-31

    Some results are given of investigations on joint operation modes of the linear induction electron accelerator LIA-30 ({approx} 40 MeV, {approx} 100 kA, {approx} 20 ns) and the pulsed reactor FR-1 with a compact metal core, aimed at achieving high intensity neutron fluxes. The multiplication factor Q for prompt neutrons in the FR-1 booster mode operation increased from 100 to 4500. The total output of prompt neutrons from FR-1 at Q = 2570 was 1.4 x 10{sup 16} 1/pulse with a pulse half width of {approx} 25 {mu}s. (author). 4 figs., 4 refs.

  9. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  10. Operator training facilities for CEGB advanced gas cooled reactors

    International Nuclear Information System (INIS)

    Green, J.F.; Birnie, S.

    1980-01-01

    The facilities provided at the Nuclear Power Training Centre of the CEGB for the training of operators fo the AGR are described. The simulator control desks are replicas of three AGR designs with, in addition, simulation of the Data Processing System for each station. Three modes of operation are envisaged: a.) Demonstration where the simulator is used by the tutor to illustrate lecture on plant behaviour. b.) Interaction where the student carries out normal procedures and experiences plant failure situations. c.) Investigation where engineering staff use the simulator for validation of modified operational procedures, ergonomic studies etc. (orig./HP)

  11. A reverse time of flight analyzer facility at the ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maayouf, R M.A.; El-Shafey, A S; Khalil, M I [Reactor and Neutron Physics Dept., NRC, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The present work deals both with the theory and performance of a reverse-time-of-flight (RTOF) analyzer designed to analyze pulses emitted from a fourier chopper recently put into operation at the ETRR-1 reactor. The RTOF analyze was found to be adequate for use with pick up pulses from the fourier chopper which operates following a frequency window suitable for rotation rates from 0-9000 rpm; synchronically with neutron pulses from a {sup 6} Li glass detector set at time focusing geometry for scattering angle 20=90 degree. It was possible, with the present RTOF analyzer to obtain diffraction patterns at neutron wavelength range between 1 - 4 A within a resolution = 0.5%. 8 FIGS.

  12. LASTRON - Second generation accelerators and chemical reactors for EBFGT facilities

    International Nuclear Information System (INIS)

    Edinger, R.

    2011-01-01

    Commercializing reliable affordable electron beam flue gas treatment technology requires both, the optimization of accelerator technology and chemical reaction chambers. Moreover, this engineering process involves the integration of beam specific characteristics, such as dosage distribution and penetration of electrons into the flue gas stream. In consideration of the treatment economy, it might be required to calculate the overall process performance without merely limiting the evaluation to accelerator efficiency. For example, a higher energy beam, 1MeV to 2 MeV, reduces the losses in the beam window and penetrates further into the gas stream and, therefore, increases the overall process economy. The energy distribution should be optimized with respect to the configuration of the chemical reaction chamber in order to treat the flue gas uniformly. All these measures are required to achieve high removal rates in large flue gas streams. Today removal rates of more than 99% SO x and more than 80% SO x are required to be compliant with future emission legislations. It is planed to establish a 100,000m³ electron beam flue gas treatment facility that can achieve constant removal rates of higher than 99.4% SO x and more than 80% NO x . The high removal rates would allow us to place CO 2 capture technologies down stream of the EBFGT facility. (author)

  13. LASTRON - Second generation accelerators and chemical reactors for EBFGT facilities

    Energy Technology Data Exchange (ETDEWEB)

    Edinger, R. [PAVAC Industries Inc. Richmond, British Columbia (Canada)

    2011-07-01

    Commercializing reliable affordable electron beam flue gas treatment technology requires both, the optimization of accelerator technology and chemical reaction chambers. Moreover, this engineering process involves the integration of beam specific characteristics, such as dosage distribution and penetration of electrons into the flue gas stream. In consideration of the treatment economy, it might be required to calculate the overall process performance without merely limiting the evaluation to accelerator efficiency. For example, a higher energy beam, 1MeV to 2 MeV, reduces the losses in the beam window and penetrates further into the gas stream and, therefore, increases the overall process economy. The energy distribution should be optimized with respect to the configuration of the chemical reaction chamber in order to treat the flue gas uniformly. All these measures are required to achieve high removal rates in large flue gas streams. Today removal rates of more than 99% SO{sub x} and more than 80% SO{sub x} are required to be compliant with future emission legislations. It is planed to establish a 100,000m³ electron beam flue gas treatment facility that can achieve constant removal rates of higher than 99.4% SO{sub x} and more than 80% NO{sub x}. The high removal rates would allow us to place CO{sub 2} capture technologies down stream of the EBFGT facility. (author)

  14. Instrument intercomparison in the pulsed neutron fieldsat the CERN HiRadMat facility

    CERN Document Server

    Aza, E; Cassell, C; Charitonidis, N; Harrouch, E; Manessi, G P; Pangallo, M; Perrin, D; Samara, E; Silari, M

    2014-01-01

    An intercomparison of the performances of active neutron detectors was carried out in pulsed neutron fi elds in the new HiRadMat facility at CERN. Five detectors were employed: four of them (two ionization chambers and two rem counters) are routinely employed in the CERN radiation monitoring system, while the fi fth is a novel instrument, called LUPIN, speci fi cally conceived for applications in pulsed neutron fi elds. The measurements were performed in the stray fi eld generated by a proton beam of very short duration with momentum of 440 GeV/c impinging on a dump. The beam intensity was steadily increased during the experiment by more than three orders of magnitude, with an H*(10) due to neutrons at the detector reference positions varying between a few nSv per burst and a few m Sv per burst, whereas the gamma contribution to the total H*(10) was negligible. The aim of the experiment was to evaluate the linearity of the detector response in extreme pulsed conditions as a function of the neutron burst in- t...

  15. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

    2006-01-01

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  16. Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

    The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

  17. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    Energy Technology Data Exchange (ETDEWEB)

    Benson, Jeff B; Marshall, Frances M [Idaho National Laboratory, Idaho Falls, ID (United States); Allen, Todd R [Univ. of Wisconsin, Madison, WI (United States)

    2012-03-15

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  18. Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

    International Nuclear Information System (INIS)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987

  19. Very high temperature gas-cooled reactor critical facility for Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Ishihara, Noriyuki

    1985-01-01

    The outline of the critical facility, its construction, the results of the basic studies and experiments on the graphite material, and the results obtained from the test conducted on the overall functions of the critical facility were reported. With the completion of the critical facility, it has been made possible to demonstrate the establishment of the manufacturing techniques and product-quality guarantee for extremely pure isotropic graphite in addition to the reliability of the structural design and analytical techniques for the main unit of the critical facility. It is expected that the present facility will prove instrumental in the verification of the nuclear safety of the very high temperature gas-cooled nuclear reactor and in the acquisition of experimental data on the reactor physics pertaining to the improvement of the reactor characteristics. The tasks which remain to be accomplished hereafter are the improvements of the performance and quality features with regard to the oxidization of graphite, the heat-resisting structural materials, and the welded structures. (Kubozono, M.)

  20. The FR 2 reactor at Karlsruhe, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FR 2 reactor and associated hot cell facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  1. Radiation-driven hydrodynamics of long pulse hohlraums on the National Ignition Facility

    International Nuclear Information System (INIS)

    Dewald, D L; Landen, O L; Suter, L J; Schein, J; Holder, J.; Campbell, K.; Glenzer, S H.; McDonald, J W.; Niemann, C.; Mackinnon, A J.; Schneider, M S.; Haynam, C.; Hinkel, D.; Hammel, B.A.

    2005-01-01

    The first hohlraum experiments on the National Ignition Facility (NIF) using the first four laser beams have activated the indirect drive experimental capabilities and tested radiation temperature limits imposed by hohlraum plasma filling. Vacuum hohlraums have been irradiated with laser powers up to 6 TW, 1 ns to 9 ns long square pulses and energies of up to 17 kJ to activate several diagnostics, to study the hohlraum radiation temperature scaling with the laser power and hohlraum size, and to make contact with hohlraum experiments performed at the NOVA and Omega laser facilities. Furthermore, for a variety of hohlraum sizes and pulse lengths, the measured x-ray flux shows signatures of plasma filling that coincide with hard x-ray emission from plasma streaming out of the hohlraum. These observations agree with hydrodynamic simulations and with analytical modeling that includes hydrodynamic and coronal radiative losses. The modeling predicts radiation temperature limits on full NIF (1.8 MJ) that are significantly greater than required for ignition hohlraums

  2. In-pile experiments and test facilities proposed for fast reactor safety

    International Nuclear Information System (INIS)

    Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

    1976-01-01

    The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

  3. New developments of TOF neutron diffraction at the IBR-2 pulsed reactor

    International Nuclear Information System (INIS)

    Balagurov, Anatoli M.

    2001-01-01

    Development of high-resolution RTOF Fourier technique for powder neutron diffraction studies is being continued at the IBR-2 pulsed reactor in Dubna. Besides some technical improvements in the operating HRFD instrument, a new dedicated instrument, Fourier Strain Diffractometer (FSD), for investigation of residual stresses in bulk materials has been constructed at IBR-2 in 1999. With a new HRFD Fourier chopper smaller than 10 μs TOF contribution in a resolution function was obtained in the experiment with perfect Si single crystal. A series of diffraction experiments with the beams from a new methane cold neutron moderator installed at the IBR-2 in 1999 is discussed. A comparison with the results obtained with the conventional water comb-like moderator shows that for various types of experiments, which are performed at HRFD and DN-2 diffractometers, the methane cold neutron source provides better conditions. (author)

  4. Neutron diffraction potentialities at the IBR-2 pulsed reactor for nondestructive testing of structural materials

    International Nuclear Information System (INIS)

    Balagurov, A.M.; Bokuchava, G.D.; Papushkin, I.V.; Sumin, V.V.; Venter, A.M.

    2010-01-01

    Neutron diffraction is widely used for investigations of residual and applied stresses in bulk materials and components. The most important factor in these investigations is the high penetration depth of neutrons (up to 2 cm for steel). At the IBR-2 pulsed reactor in Dubna the Fourier stress diffractometer (FSD) has been constructed to optimize the internal stress measurements. The FSD design satisfies the requirements of high luminosity, high resolution and specific sample environment. The collimator system guarantees a minimum gauge volume of 2x2x2 mm. A mechanical testing machine allows in-situ tension or compression measurements up to a load of 20 kN and sample temperatures up to 800 deg C. In the paper the current status of FSD is reported and potentialities are demonstrated with several examples of investigations performed

  5. Heat transfer in the lithium-cooled blanket of a pulsed fusion reactor

    International Nuclear Information System (INIS)

    Cort, G.E.; Krakowski, R.A.

    1978-01-01

    The transient temperature distribution in the lithium-cooled blanket of a pulsed fusion reactor has been calculated using a finite-element heat-conduction computer program. An auxiliary program was used to predict the coolant transient velocity in a network of parallel and series flow passages with constant driving pressure and varying magnetic field. The coolant velocity was calculated by a Runge-Kutta numerical integration of the conservation equations. The lithium coolant was part of the finite-element heat-conduction mesh with the velocity terms included in the total matrix. The matrix was solved implicitly at each time step for the nodal point temperatures. Slug flow was assumed in the coolant passages and the Boussinesq analogy was used to calculate turbulent heat transfer when the magnetic field was not present

  6. Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1997-11-01

    A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  7. 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

  8. Description of the PIE facility for research reactors irradiated fuels in CNEA

    International Nuclear Information System (INIS)

    Bisca, A.; Coronel, R.; Homberger, V.; Quinteros, A.; Ratner, M.

    2002-01-01

    The PIE Facility (LAPEP), located at the Ezeiza Atomic Center (CAE), was designed to carry out destructive and non-destructive post-irradiation examinations (PIE) on research and power reactor spent fuels, reactor internals and other irradiated materials, and to perform studies related with: Station lifetime extension; Fuel performance; Development of new fuels; and Failures and determination of their causes. LAPEP is a relevant facility where research and development can be carried out. It is worth mentioning that in this facility the PIE corresponding to the Surveillance Program for the Atucha I Nuclear Power Plant (CNA-1) were successfully performed. Materials testing during the CNA-1 repair and the study of failures in fuel element plugs of the Embalse Nuclear Power Plant (CNE) were also performed. (author)

  9. Preliminary studi on neutronic aspect of a conceptual design of the Kartini reactor base ADS facility

    International Nuclear Information System (INIS)

    Tegas Sutondo

    2012-01-01

    A preliminary study on neutronic aspect of a conceptual design of ADS facility with the basis of Kartini Reaktor, has been performed. The study was intended to see the feasibility from neutronic point of view of Kartini reactor, to be used as a small scale of NPP’s waste transmutation experimental facility. A SRAC code was used as the basis of calculations. The results indicate that the presence of minor actinides (MA) will give a positive reactivity, which tends to increase with the increase of MA concentrations. Based on the defined criteria of subcriticality and by considering the core power distributions and the level of reactivity contribution of MA element, it is concluded that Kartini reactor is potential enough to be used as an ADS experimental facility, mainly for MA concentration between 30 to 50 % of the assumed mixture of C-MA matrix. (author)

  10. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

    1983-01-01

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.) [de

  11. Conceptual design of a fission-based integrated test facility for fusion reactor components

    International Nuclear Information System (INIS)

    Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

    1982-01-01

    The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

  12. Fast-neutron dosimetry in the seed-irradiation facility, ASTRA reactor, Seibersdorf

    International Nuclear Information System (INIS)

    Ahnstroem, G.; Burtscher, A.; Casta, J.

    1967-01-01

    An important part of the co-ordinated programme on the neutron irradiation of seeds has been the construction of a fast-neutron irradiation facility for swimming-pool reactors. This facility was installed around 70 cm from the core in the ASTRA reactor swimming-pool at the end of December, 1966. Also, for this programme a pair of constant potential ionization chambers have been constructed at the Institute of Biochemistry, Stockholm University. These chambes are of the type described in the technical annex and are the same size as the seed-irradiation vials to be used in the seed-irradiation container (diam. =15 mm, length = 60 mm). Some preliminary dosimetry experiments were undertaken to test the irradiation facility and the ionization chambers, and to investigate the usefulness of the dosimetry instructions in the Technical Annex. The results of these experiments are discussed in this paper. 3 refs, 6 figs, 7 tabs

  13. Pulse Star Inertial Confinement Fusion Reactor: Heat transfer loop and balance-of-plant considerations

    International Nuclear Information System (INIS)

    McDowell, M.W.; Blink, J.A.; Curlander, K.A.

    1983-01-01

    A conceptual heat transfer loop and balance-of-plant design for the Pulse Star Inertial Confinement Fusion Reactor has been investigated and the results are presented. The Pulse Star reaction vessel, a perforated steel bell jar about11 m in diameter, is immersed in Li 17 Pb 83 coolant, which flows through the perforations and forms a 1.5-m-thick plenum of droplets around a 8-m-diameter inner chamber. The bell jar and associated pumps, piping, and steam generators are contained within a 17-m-diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops, each with a flow rate of 5.5 m 3 /s, are necessary to transfer 3300 MWt of power. Liquid metal is pumped to the top of the pool, where it flows downward through eight vertical steam generators. Double-walled tubes are used in the steam generators to assure tritium containment without intermediate heat transfer loops. Each pump is a mixed flow type and has a required NPSH of 3.4 m, a speed of 278 rpm, and an impeller diameter of 1.2 m. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. The design minimizes the total cost (heat exchanger area plus pumping) for the plant lifetime. The power required for the pumps is 36 MWe. Each resulting steam generator is 12 m high and 1.6 m in diameter, with 2360 tubes. The steam generators and pumps fit easily in the pool between the reactor chamber and the pool wall

  14. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  15. The University of Missouri Research Reactor facility can melter system

    International Nuclear Information System (INIS)

    Edwards, C.B. Jr.; Olson, O.L.; Stevens, R.; Brugger, R.M.

    1987-01-01

    At the University of Missouri Research Reactor (MURR), a waste compacting system for reducing the volume of radioactive aluminum cans has been designed, built and put into operation. In MURR's programs of producing radioisotopes and transmutation doping of silicon, a large volume of radioactive aluminum cans is generated. The Can Melter System (CMS) consists of a sorting station, a can masher, an electric furnace and a gas fired furnace. This system reduces the cans and other radioactive metal into barrels of solid metal close to theoretical density. The CMS has been in operation at the MURR now for over two years. Twelve hundred cu ft of cans and other metals have been reduced into 150 cu ft of shipable waste. The construction cost of the CMS was $4950.84 plus 1680 man hours of labor, and the operating cost of the CMS is $18/lb. The radiation exposure to the operator is 8.6 mR/cu ft. The yearly operating savings is $30,000. 20 figs., 10 tabs

  16. Guide to the safety design examination about light water reactor facilities for power generation

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This guide was compiled to evaluate the validity of the design policy when the safety design is examined at the time of the application for approval of the installation of nuclear reactors. About 7 years has elapsed since the existing guide was established, and the more appropriate guide to evaluate the safety should be made on the basis of the knowledge and experience accumulated thereafter. The range of application of this guide is limited to the above described evaluation, and it is not intended as the general standard for the design of nuclear reactors. First, the definition of the words used in this guide is given. Then, the guide to the safety examination is described about the general matters of reactor facilities, nuclear reactors and the measuring and controlling system, reactor-stopping system, reactivity-controlling system and safety protection system, reactor-cooling system, reactor containment vessels, fuel handling and waste treatment system. Several matters which require attention in the application of this guide or the clarification of the significance and interpretation of the guide itself were found, therefore the explanation about them was added at the end of this guide. (Kako, I.)

  17. 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)

  18. HiRadMat at CERN/SPS - A dedicated facility providing high intensity beam pulses to material samples

    CERN Multimedia

    Charitonidis, N; Efthymiopoulos, I

    2014-01-01

    HiRadMat (High Radiation to Materials), constructed in 2011, is a facility at CERN designed to provide high‐intensity pulsed beams to an irradiation area where material samples as well as accelerator component assemblies (e.g. vacuum windows, high power beam targets, collimators…) can be tested. The facility uses a 440 GeV proton beam extracted from the CERN SPS with a pulse length of up to 7.2 us, and with a maximum pulse energy of 3.4 MJ (3xE13 proton/pulse). In addition to protons, ion beams with energy of 440 GeV/charge and total pulse energy of 21 kJ can be provided. The beam parameters can be tuned to match the needs of each experiment. HiRadMat is not an irradiation facility where large doses on equipment can be accumulated. It is rather a test area designed to perform single pulse experiments to evaluate the effect of high‐intensity pulsed beams on materials or accelerator component assemblies in a controlled environment. The fa‐ cility is designed for a maximum of 1E16 protons per year, dist...

  19. Transition between trickle flow and pulse flow in a cocurrent gas-liquid trickle-bed reactor at elevated pressures

    NARCIS (Netherlands)

    Wammes, W.J.A.; Mechielsen, S.J.; Westerterp, K.R.

    1992-01-01

    The effect of reactor pressure in the range of 0.2–2.0 MPa on the transition between the trickle-flow and the pulse-flow regime has been investigated for the non-foaming water—nitrogen and aqueous 40% ethyleneglycol—nitrogen systems. Most models and flow charts which are all based on atmospheric

  20. The U.S. DOE new production reactor/heavy water reactor facility pollution prevention/waste minimization program

    International Nuclear Information System (INIS)

    Kaczmarsky, Myron M.; Tsang, Irving; Stepien, Walter P.

    1992-01-01

    A Pollution Prevention/Waste Minimization Program was established during the early design phase of the U.S. DOE's New Production Reactor/Heavy Water Reactor Facility (NPR/HWRF) to encompass design, construction, operation and decommissioning. The primary emphasis of the program was given to waste elimination, source reduction and/or recycling to minimize the quantity and toxicity of material before it enters the waste stream for treatment or disposal. The paper discusses the regulatory and programmatic background as it applies to the NPR/HWRF and the waste assessment program developed as a phased approach to pollution prevention/waste minimization for the NPR/HWRF. Implementation of the program will be based on various factors including life cycle cost analysis, which will include costs associated with personnel, record keeping, transportation, pollution control equipment, treatment, storage, disposal, liability, compliance and oversight. (author)

  1. Applicability of base-isolation R and D in non-reactor facilities to a nuclear reactor plant

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1989-01-01

    Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. This paper reviews the research and development (R and D) programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R and D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R and D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant

  2. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  3. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges - 15066

    International Nuclear Information System (INIS)

    Sabharwall, P.; O'Brien, J.E.; Yoon, S.J.; Sun, X.

    2015-01-01

    A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic, materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The 3 loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuits heat exchangers (PCHEs) at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integrated System Test (ARTIST) facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 C. degrees), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF 4 ) flow loop operating at low pressure (0.2 MPa), at a temperature of ∼ 450 C. degrees. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift) in measuring operational data for extended periods of times, as data collected will be

  4. Inactivation of Escherichia coli in water by pulsed dielectric barrier discharge in coaxial reactor.

    Science.gov (United States)

    Hernández-Arias, A N; Rodríguez-Méndez, B G; López-Callejas, R; Alcántara-Díaz, D; Valencia-Alvarado, R; Mercado-Cabrera, A; Peña-Eguiluz, R; Muñoz-Castro, A E; Barocio, S R; de la Piedad-Beneitez, A

    2012-09-01

    An experimental study of ATCC (American Type Culture Collection) 8739 Escherichia coli bacteria inactivation in water by means of pulsed dielectric barrier discharge (PDBD) atmospheric pressure plasmas is presented. Plasma is generated by an adjustable power source capable of supplying high voltage 25 kV pulses, ∼30 μs long and at a 500 Hz frequency. The process was conducted in a ∼152 cm(3) cylindrical stainless steel coaxial reactor, endowed with a straight central electrode and a gas inlet. The bacterial concentration in water was varied from 10(3) up to 10(8) E. coli cells per millilitre. The inactivation was achieved without gas flow in the order of 82% at 10(8) colony-forming units per millilitre (CFU mL(-1)) concentrations in 600 s. In addition, oxygen was added to the gas supply in order to increase the ozone content in the process, raising the inactivation percentage to the order of 90% in the same treatment time. In order to reach a higher efficiency however, oxygen injection modulation is applied, leading to inactivation percentages above 99.99%. These results are similarly valid for lower bacterial concentrations.

  5. 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.

  6. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  7. Report of specialists' meeting on 'pulse reactor using the particle accelerators'

    International Nuclear Information System (INIS)

    Inoue, Makoto; Yoshiie, Toshimasa

    1996-06-01

    KUR was decided to be continued to operate at the Reactor Laboratory according to the report on 'How to investigate the reactor for study in the Universities' published by Committee of Science Consideration on July, 1997. However, it is necessary to proceed the next program based on the KUR study results in future. From a viewpoint of the study, on considering of beginning of the next program during using KUR still but not establishing it after wasting the KUR, it is preferable to be an equipment with complemental characters for the KUR. As the pulse neutron source, spallation neutron source using large intensity proton accelerator with about 1 GeV is focused most of interests. Use of the proton beam with large neutron generation in comparison with electron beam can make non-critical collector with relatively large non-critical ratio to a target, and can obtain safety without any change of reactivity by means of mechanical method. Furthermore, this accelerator has some times of the non-critical ratio in comparison with the spallation method using only accelerator and has a feature to be low in its cost. In this report, its program, 11 items of lectures and general discussion on them were described. (G.K.)

  8. Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

    PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

  9. A midsize reactor facility - A regional resource for research and education

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1991-01-01

    The mission of the University of Florida Training Reactor (UFTR) is to serve the regional needs of Florida and the Southeast for access to quality reactor usage. Well-advertised capabilities of the facility support diversified usages that include education, training, research, service, and public information programs to address the needs of a broad spectrum of users ranging from high school students and teachers, to university researchers, and even the occasional service user. Despite the midsize power of the facility, the UFTR's status as the only nonpower reactor within 350 miles in one of our largest states means that it is uniquely situated to contribute in these various areas in ways usually reserved for larger facilities. Nine state universities and a well-developed community college system in addition to private schools and a growing complement of progressive high schools assure a broad-based user community. The key to accomplishing mission objectives is to continue diversification and improvement of both the reactor and associated experimental capabilities to meet the needs of this user community

  10. TIT reactor laboratory course using JAERI and PNC large experimental facilities

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Obara, Toru; Ohtani, Nobuo.

    1995-01-01

    This report is presented on a reactor laboratory course for graduate students using large facilities in national laboratories in Japan. A reactor laboratory course is offered every summer since 1990 for all graduate students in the Nuclear Engineering Course in Tokyo Institute of Technology (TIT), where the students can choose one of the experiments prepared at Japan Atomic Energy Research Institute (JAERI), Power Reactor and Nuclear Fuel Development Corporation (PNC) and Research Reactor Institute, Kyoto University (KUR). Both JAERI and PNC belong to Science and Technology Agency (STA). This is the first university curriculum of nuclear engineering using the facilities owned by the STA laboratories. This type of collaboration is promoted in the new Long-Term Program for Research, Development and Utilization of Nuclear Energy adopted by Atomic Energy Commission. Most students taking this course reported that they could learn so much about reactor physics and engineering in this course and the experiment done in large laboratory was a very good experience for them. (author)

  11. Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Nystrand, A.C.; Lassing, A.

    1999-01-01

    Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

  12. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bissani, M; O' Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

  13. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® Reactor

    International Nuclear Information System (INIS)

    Schickler, R.A.; Marcum, W.R.; Reese, S.R.

    2013-01-01

    Highlights: • The Oregon State TRIGA ® Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA ® Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least-squares technique. The quantification of

  14. Modernization of the facilities of the TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Mendez T, D.; Flores C, J.

    2016-09-01

    The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

  15. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Malkoske, G.R.

    2003-01-01

    The availability of isotope grade, Highly Enriched Uranium (HEU), from the United States for use in the manufacture of targets for molybdenum-99 production in AECL's NRU research reactor has been a key factor to enable MDS Nordion to develop a reliable, secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets is a proven and established method that has reliably produced medical isotopes for several decades. The HEU process provides predictable, consistent yields for our high-volume, molybdenum-99 production. Other medical isotopes such as I-131 and Xe-133, which play an important role in nuclear medicine applications, are also produced from irradiated HEU targets as a by-product of the molybdenum-99 process. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the commissioning of two MAPLE reactors and an associated isotope processing facility (the New Processing Facility). The new MAPLE facilities, which will be dedicated exclusively to medical isotope production, will provide an essential contribution to a secure, robust global healthcare system. Design and construction of these facilities has been based on a life cycle management philosophy for the isotope production process. This includes target irradiation, isotope extraction and waste management. The MAPLE reactors will operate with Low Enriched Uranium (LEU) fuel, a significant contribution to the objectives of the RERTR program. The design of the isotope production process in the MAPLE facilities is based on an established process - extraction of isotopes from HEU target material. This is a proven technology that has been demonstrated over more than three decades of operation. However, in support of the RERTR program and in compliance with U.S. legislation, MDS Nordion has undertaken a LEU Target Development and Conversion Program for the MAPLE facilities. This paper will provide an

  16. Design of Safety Parameter Monitoring Function in a Research Reactor Facility

    International Nuclear Information System (INIS)

    Park, Jaekwan; Suh, Yongsuk

    2014-01-01

    The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

  17. Design of Safety Parameter Monitoring Function in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

  18. Use Of Cementitious Materials For SRS Reactor Facility In-Situ Decommissioning - 11620

    International Nuclear Information System (INIS)

    Langton, C.; Stefanko, D.; Serrato, M.; Blankenship, J.; Griffin, W.; Waymer, J.; Matheny, D.; Singh, D.

    2010-01-01

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD requires about 250,000 cubic yards of grout to fill the below grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Cementitious materials were designed for the following applications: (1) Below grade massive voids/rooms: Portland cement-based structural flowable fills for - Bulk filling, Restricted placement and Underwater placement. (2) Special below grade applications for reduced load bearing capacity needs: Cellular portland cement lightweight fill (3) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels: Calcium sulfoaluminate flowable fill, and Magnesium potassium phosphate flowable fill. (4) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2012. The focus of this paper is to describe the (1) grout mixes

  19. USE OF CEMENTITIOUS MATERIALS FOR SRS REACTOR FACILITY IN-SITU DECOMMISSIONING - 11620

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Stefanko, D.; Serrato, M.; Blankenship, J.; Griffin, W.; Waymer, J.; Matheny, D.; Singh, D.

    2010-12-07

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD requires about 250,000 cubic yards of grout to fill the below grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Cementitious materials were designed for the following applications: (1) Below grade massive voids/rooms: Portland cement-based structural flowable fills for - Bulk filling, Restricted placement and Underwater placement. (2) Special below grade applications for reduced load bearing capacity needs: Cellular portland cement lightweight fill (3) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels: Calcium sulfoaluminate flowable fill, and Magnesium potassium phosphate flowable fill. (4) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2012. The focus of this paper is to describe the (1) grout mixes

  20. Scram and nonlinear reactor system seismic analysis for the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Morrone, A.

    1975-01-01

    A description is given of the analysis and results for the Fast Flux Test Facility (FFTF) reactor system which was analyzed for both scram times and seismic responses such as bending moments and impact forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node. The results give time history plots of various seismic responses and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. (U.S.)

  1. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits

  2. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.

  3. Reports and operational engineering: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Rochman, A.; Washburn, B.W.

    1981-02-01

    The Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, established via an October 24, 1979 memorandum from the Department of Energy (DOE) Under Secretary, was instructed to review the ''Kemeny Commission'' recommendations and to identify possible implications for DOE's nuclear facilities. As a result of this review, the Committee recommended that DOE carry out assessments in seven categories. The assessments would address specific topics identified for each category as delineated in the NFPQT ''Guidelines for Assessing the Safe Operation of DOE-Owned Reactors,'' dated May 7, 1980. The Committee recognized that similar assessments had been ongoing in the DOE program and safety overview organizations since the Three Mile Island nuclear accident and it was the Committee's intent to use the results of those ongoing assessments as an input to their evaluations. This information would be supplemented by additional studies consisting of the subject-related documents used at each reactor facility studied, and an on-site review of these reactor facilities by professional personnel within the Department of Energy, its operating contractors and independent consultants. 1 tab

  4. Detailed description of an SSAC at the facility level for light water moderated (off-load refueled) power reactor facilities

    International Nuclear Information System (INIS)

    Jones, R.J.

    1985-03-01

    This report is intended to provide the technical details of an effective State Systems of Accounting for and Control of Nuclear Material (SSAC) which Member States may use, if they wish, to establish and maintain their SSACs. It is expected that systems designed along the lines described would be effective in meeting the objectives of both national and international systems for nuclear material accounting and control. This document accordingly provides a detailed description of a system for the accounting for and control of nuclear material in an off-load refueled light water moderated power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  5. Waste from decommissioning of research reactors and other small nuclear facilities

    International Nuclear Information System (INIS)

    Massaut, V.

    2001-01-01

    Full text: Small nuclear facilities were often built for research or pilot purposes. It includes the research reactors of various types and various aims (physics research, nuclear research, nuclear weapons development, materials testing reactor, isotope production, pilot plant, etc.) as well as laboratories, hot cells and accelerators used for a broad spectrum of research or production purposes. These installations are characterized not only by their size (reduced footprint) but also, and even mostly, by the very diversified type of materials, products and isotopes handled within these facilities. This large variety can sometimes enhance the difficulties encountered for the dismantling of such facilities. The presence of materials like beryllium, graphite, lead, PCBs, sodium, sometimes in relatively large quantities, are also challenges to be faced by the dismantlers of such facilities, because these types of waste are either toxic or no solutions are readily available for their conditioning or long term disposal. The paper will review what is currently done in different small nuclear facilities, and what are the remaining problems and challenges for future dismantling and waste management. The question of whether Research and Development for waste handling methods and processes is needed is still pending. Even for the dismantling operation itself, important improvements can be brought in the fields of characterization, decontamination, remote handling, etc. by further developments and innovative systems. The way of funding such facilities decommissioning will be reviewed as well as the very difficult cost estimation for such facilities, often one-of-a-kind. The aspects of radioprotection optimization (ALARA principle) and classical operators safety will also be highlighted, as well as the potential solutions or improvements. In fact, small nuclear facilities encounter often, when dismantling, the same problems as the large nuclear power plants, but have in

  6. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

  7. Use of Cementitious Materials for SRS Reactor Facility In-Situ Decommissioning

    International Nuclear Information System (INIS)

    Langton, C.A.; Stefanko, D.B.; Serrato, M.G.; Blankenship, J.K.; Griffin, W.G.; Long, J.T.

    2013-01-01

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose of producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD project requires approximately 250000 cubic yards of cementitious materials to fill the below-grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Funding is being provided under the American Recovery and Reinvestment Act (ARRA). Cementitious materials were designed for the following applications: (A) Below-grade massive voids / rooms: Portland cement-based structural flowable fills for: (A.1) Bulk filling; (A.2) Restricted placement and (A.3) Underwater placement. (B) Special below-grade applications for reduced load bearing capacity needs: (B.1) Cellular portland cement lightweight fill. (C) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels (C.1) Blended calcium aluminate - calcium sulfate based flowable fill; (C.2) Magnesium potassium phosphate flowable fill. (D) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: (D.1) Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured

  8. 77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

    Science.gov (United States)

    2012-11-15

    ... Research Institute TRIGA Reactor: Facility Operating License No. R-84 AGENCY: Nuclear Regulatory Commission... considering an application for the renewal of Facility Operating License No. R-84 (Application), which... the renewal of Facility Operating License No. R-84, which currently authorizes the licensee to operate...

  9. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  10. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  11. Applicability of base-isolation R ampersand D in non-reactor facilities to a nuclear reactor plant

    International Nuclear Information System (INIS)

    Seidensticker, R.W.; Chang, Y.W.

    1990-01-01

    Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R ampersand D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R ampersand D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R ampersand D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R ampersand D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs

  12. A description of the demonstration Integral Fast Reactor fuel cycle facility

    International Nuclear Information System (INIS)

    Courtney, J.C.; Carnes, M.D.; Dwight, C.C.; Forrester, R.J.

    1991-01-01

    A fuel examination facility at the Idaho National Engineering Laboratory is being converted into a facility that will electrochemically process spent fuel. This is an important step in the demonstration of the Integral Fast Reactor concept being developed by Argonne National Laboratory. Renovations are designed to bring the facility up to current health and safety and environmental standards and to support its new mission. Improvements include the addition of high-reliability earthquake hardened off-gas and electrical power systems, the upgrading of radiological instrumentation, and the incorporation of advances in contamination control. A major task is the construction of a new equipment repair and decontamination facility in the basement of the building to support operations

  13. CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Haden, T.H.J.J. van der; Zboray, R.; Manera, A.; Mudde, R.F.

    2002-01-01

    At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

  14. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Thelen, M.C.; Meyer, M.K.; Marshall, F.M.; Foster, J.; Benson, J.B.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  15. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Benson, J.B.; Foster, J.A.; Marshall, F.M.; Meyer, M.K.; Thelen, M.C.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  16. Independent Confirmatory Survey Summary and Results for the Plum Brook Reactor Facility Sandusky OH

    International Nuclear Information System (INIS)

    Bailey, E.N.

    2008-01-01

    In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics (later known as the National Aeronautics and Space Administration or NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventually built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities. After successfully completing the objective of landing humans on the Moon and returning them safely to Earth, NASA was faced with budget reductions from Congress in 1973. These budgetary constraints caused NASA to cease operations at several research facilities across the country, including those at Plum Brook Station. The major test facilities at Plum Brook were maintained in a standby mode, capable of being reactivated for future use. The Plum Brook Reactor Facility (PBRF) was shut down January 5, 1973 and all of the nuclear fuel was eventually removed and shipped off site to a U.S. Department of Energy facility in Idaho for disposal or reuse. Decommissioning activities are currently underway at the PBRF (NASA 1999). The objectives of the confirmatory survey activities were to provide independent contractor field data reviews and to generate independent radiological data for use by the Nuclear Regulatory Commission (NRC) in evaluating the adequacy and accuracy of the licensee's procedures and final status survey (FSS) results

  17. Research and design of pulsed switching power supply for deep tumor therapy facility with heavy ions accelerator in Lanzhou

    International Nuclear Information System (INIS)

    Shi Chunfeng; Zhao Jiang; Yan Hongbin; Wu Fengjun; Gao Daqing

    2012-01-01

    The pulsed switching power supply was developed for deep tumor therapy facility with heavy ions in cooler-storage-ring of the heavy ions research facility in Lanzhou (HIRFL-CSR). The control principle of the dual closed-loop scheme was described and the open-loop Bode diagrams were given. The results of simulation and prototype experiment show that the current error gets much smaller than that of the single closed-loop pulsed switching power supply. Moreover, the simulation and test results were analyzed, and the circuit configuration and dual closed-loop strategy selected are practicable. (authors)

  18. Safety report content and development for test loop facility on MARIA reactor

    International Nuclear Information System (INIS)

    Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

    1982-01-01

    A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

  19. Characterization and adjustment of the neutron radiography facility of the RP-10 nuclear reactor

    International Nuclear Information System (INIS)

    Ravello R, Y.R.

    2001-01-01

    The main aim of this work was to characterize and adjust the neutron radiography facility of the RP-10 nuclear reactor, and therefore be able to offer with this technique services to the industry and research centers in general. This technique will be complemented with others such as x-rays and gamma radiography. First, the shielding capacity of the facility was analyzed, proving that it complies with the radiological safety requirements established by the radiological safety code. Then gamma filtration tests were conducted in order to implement the direct method for image formation, optical density curves were built according to the thickness of the gamma filter, the type of film and the type of irradiation. Also, the indirect method for image formation was implemented for two types of converters: indium and dysprosium. Growth curves for optical density were also made according to contact time between converter-film, for different types of films. The resolution of the facility was also analyzed using two methods: Klasens (1946) and Harms (1986). Harms method came closer to the resolution of the human eye when compared to the Klasens method. Finally, the application fields of neutron radiography are presented, including those conducted at the neutron radiography facility of the RP-10 nuclear reactor. With this work, the RP-10 neutron radiography facility is ready to offer inspection and research services

  20. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Malkoske, G.R.

    2002-01-01

    Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada has been extracted from reactor targets employing highly enriched uranium (HEU). A reliable supply of HEU metal from the United States used in the manufacture of targets for the NRU research reactor has been a key factor to enable MDS Nordion to develop a secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets provides predictable, consistent yields for our high-volume molybdenum production process. Each link of the isotope supply chain, from isotope production to ultimate use by the physician, has been established using this proven and established method of HEU target irradiation and processing to extract molybdenum-99. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the construction of two MAPLE reactors and a New Processing Facility. The design of the MAPLE facilities was based on an established process developed by Atomic Energy of Canada Ltd. (AECL) - extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a three phase LEU Target Development and Conversion Program for the MAPLE facilities. Phase 1, the Initial Feasibility Study, which identified the technical issues to convert the MAPLE reactor targets from HEU to LEU for large scale commercial production was reported on at the RERTR- 2000 conference. The second phase of the LEU Target Development and Conversion Program was developed with extensive consultation and involvement of experts knowledgeable in target development, process system design, enriched uranium conversion chemistry and commercial scale reactor operations and molybdenum production. This paper will provide an overview of the Phase 2 Conversion Development Program, report on progress to date, and further

  1. Facility at CIRUS reactor for thermal neutron induced prompt γ-ray spectroscopic studies

    International Nuclear Information System (INIS)

    Biswas, D.C.; Danu, L.S.; Mukhopadhyay, S.; Kinage, L.A.; Prashanth, P.N.; Goswami, A.; Sahu, A.K.; Shaikh, A.M.; Chatterjee, A.; Choudhury, R.K.; Kailas, S.

    2013-01-01

    A facility for prompt γ-ray spectroscopic studies using thermal neutrons from a radial beam line of Canada India Research Utility Services (CIRUS) reactor, Bhabha Atomic Research Centre (BARC), has been developed. To carry out on-line spectroscopy experiments, two clover germanium detectors were used for the measurement of prompt γ rays. For the first time, the prompt γ–γ coincidence technique has been used to study the thermal neutron induced fission fragment spectroscopy (FFS) in 235 U(n th , f). Using this facility, experiments have also been carried out for on-line γ-ray spectroscopic studies in 113 Cd(n th , γ) reaction

  2. On fire risk/methodology for the next generation of reactors and nuclear facilities

    International Nuclear Information System (INIS)

    Majumdar, K.C.; Alesso, H.P.; Altenbach, T.J.

    1992-01-01

    Methodologies for including fire in probabilistic risk assessments (PRAs) have been evolving during the last ten years. Many of these studies show that fire risk constitutes a significant percentage of external events, as well as the total core damage frequency. This paper summarizes the methodologies used in the fire risk analysis of the next generation of reactors and existing DOE nuclear facilities. Methodologies used in other industries, as well as existing nuclear power plants, are also discussed. Results of fire risk studies for various nuclear plants and facilities are shown and compared

  3. The small-angle neutron scattering facility at the SAFARI-1 reactor

    International Nuclear Information System (INIS)

    Hofmeyr, C.; Mayer, R.M.; Tillwick, D.L.

    1979-01-01

    A long-wavelength neutron scattering facility at the SAFARI-1 reactor is described. Neutrons of wavelength between 5 and 15 A can be selected. Features of the facility are the use of microwave guides as neutron conductors, flexible guide-pipe configuration and automatic sequential sample changing. Examples are given of measurements on radiation-induced voids in copper, aluminium, Al-0.4%Si and Al-0.1%In after neutron irradiation and magnetic scattering in US and in 80US-10UC-10UC 2 . (Auth.)

  4. Design of neutron radiography facility in pool for the reactor RA-10

    International Nuclear Information System (INIS)

    Peirone, M.; Coleff, A.; Sanchez, F.; Chiaraviglio, N.

    2013-01-01

    RA-10 project consists in the design and construction of a multipurpose reactor for multiple applications, including radioisotopes production, material testing and an in pool facility for neutron imaging. Neutron imaging is a powerful tool for studies of materials and offer several advantages among other attenuation-based techniques. In this study mechanical and neutronic requirements for the RA-10 in pool neutron imaging facility are described. The MCNP neutronic model and the mechanical design satisfying these requirements in a first engineering stage are described. (author)

  5. An overview of FFTF [Fast Flux Test Facility] contributions to Liquid Metal Reactor Safety

    International Nuclear Information System (INIS)

    Waltar, A.E.; Padilla, A. Jr.

    1990-11-01

    The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs

  6. Detailed description of an SSAC at the facility level for research reactors

    International Nuclear Information System (INIS)

    Jones, R.J.

    1984-09-01

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in a research reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  7. Opinions of the well-informed persons about the nuclear reactor facility periodical inspection

    International Nuclear Information System (INIS)

    Aeba, Yoichi; Ishikawa, Michio; Enomoto, Toshiaki; Oomori, Katsuyoshi

    2005-01-01

    Falsifications of self-inspection records in the shrouds and of leakage rates for containment vessels at TEPCO nuclear power plants destroyed public trust in nuclear safety. The Nuclear Reactor Regulation Law and Electric Utility Law were amended to enhance the nuclear safety regulation system. The major improvements are that operators are legally required to conduct inspection (periodical operator inspection) and recording and keeping inspection results. The operator performs 'periodical operator inspection' regularly, and Nuclear and Industrial Safety Agency (NISA) performs periodical inspection' about particularly important facilities/function in safety. Sixteen opinions of well-informed persons about the nuclear reactor facility periodical inspection were presented in this special number. Interval of periodical inspection less than thirteen months was disputed. Maintenance activities should be more rationalized based on risk information. (T. Tanaka)

  8. Decontamination and decommissioning of the Organic Moderated Reactor Experiment facility (OMRE)

    International Nuclear Information System (INIS)

    Hine, R.E.

    1980-09-01

    This report describes the decontamination and decommissioning (D and D) of the Organic Moderated Reactor Experiment (OMRE) facility performed from October 1977 through September 1979. This D and D project included removal of all the facilities and as much contaminated soil and rock as practical. Removal of the reactor pressure vessel was an unusually difficult problem, and an extraordinary, unexpected amount of activated rock and soil was removed. After removal of all significantly contaminated material, the site consisted of a 20-ft deep excavation surrounded by backfill material. Before this excavation was backfilled, it and the backfill material were radiologically surveyed and detailed records made of these surveys. After the excavation was backfilled and graded, the site surface was surveyed again and found to be essentially uncontaminated

  9. Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Israr, M.; Shami, Qamar-ud-din; Pervez, S.

    1997-11-01

    In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

  10. Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

    International Nuclear Information System (INIS)

    Lear, G.; Erickson, P.B.

    1975-01-01

    Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

  11. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  12. Radiological considerations of the reactor cover gas processing system at the FFTF [Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Prevo, P.R.

    1986-09-01

    Radiological and environmental protection experience associated with the reactor cover gas processing system at the Fast Flux Test Facility (FFTF) has been excellent. Personnel radiation exposures received from operating and maintaining the reactor cover gas processing system have been very low, the system has remained free of radioactive particulate contamination through the first seven operating cycles (cesium contamination was detected at the end of Cycle 8A), and releases of radioactivity to the environment have been very low, well below environmental standards. This report discusses these three aspects of fast reactor cover gas purification over the first eight operating cycles of the FFTF (a duration of a little more than four years, from April 1982 through July 1986)

  13. Conceptual design of PF coil system and operation scenario on inductively-operated day-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

    Wang, J.F.; Yamamoto, T.; Ogawa, Y.

    1994-01-01

    It is said that disadvantages of pulsed operation in tokamak fusion reactor are fatigue problem of structural materials and an introduction of energy storage System to compensate the power during the dwell time. To overcome theses disadvantages the authors have designed an inductively-operated ultralong pulsed tokamak called (IDLT) reactor where plasma with a major radius of 10 m are employed so as to provide a magnetic flux necessary to sustain a plasma current inductively during 10 hours or more. This makes it possible to reduce the total cycle number to be around 10 4 during the life of the fusion plant. In pulsed operation reactors the shorter dwell time with a quick start-up and shut down of plasma is very convenient to realize a high availability of the power plant, but it will induce more severe conditions for the hardware design. The authors assumed the dwell time of 5∼10 minutes and analyzed the feasibility of plasma operation scenario for IDLT reactor, especially paying much attention to PF coil system. The stored energy of PF coil system becomes ∼100 GJ, which is comparable with that of toroidal field coil system. When the plasma current of 14 MA is ramped up with a time of 100 seconds, it is found that the maximum capacity of 1 GW is necessary for PF coil power supply. Engineering issues related with AC/hysterisis loss should be carefully examined

  14. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bertini, H.W.

    1980-04-01

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

  15. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    International Nuclear Information System (INIS)

    Bertini, H.W.

    1980-04-01

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established

  16. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  17. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1995-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy's Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period

  18. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.

  19. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs

  20. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    Roettger, H.; Hardt, P. von der; Tas, A.; Voorbraak, W.P.

    1981-01-01

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to January 1981

  1. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  2. Development of a Test Facility to Simulate the Reactor Flow Distribution of APR+

    International Nuclear Information System (INIS)

    Euh, D. J.; Cho, S.; Youn, Y. J.; Kim, J. T.; Kang, H. S.; Kwon, T. S.

    2011-01-01

    Recently a design of new reactor, APR+, is being developed, as an advanced type of APR1400. In order to analyze the thermal margin and hydraulic characteristics of APR+, quantification tests for flow and pressure distribution with a conservation of flow geometry are necessary. Hetsroni (1967) proposed four principal parameters for a hydraulic model representing a nuclear reactor prototype: geometry, relative roughness, Reynolds number, and Euler number. He concluded that the Euler number should be similar in the prototype and model under the preservation of the aspect ratio on the flow path. The effect of the Reynolds number at its higher values on the Euler number is rather small, since the dependency of the form and frictional loss coefficients on the Reynolds number is seen to be small. ABB-CE has carried out several reactor flow model test programs, mostly for its prototype reactors. A series of tests were conducted using a 3/16 scale reactor model. (see Lee et al., 2001). Lee et al (1991) performed experimental studies using a 1/5.03 scale reactor flow model of Yonggwang nuclear units 3 and 4. They showed that the measured data met the acceptance criteria and were suitable for their intended use in terms of performance and safety analyses. The design of current test facility was based on the conservation of Euler number which is a ratio of pressure drop to dynamic pressure with a sufficiently turbulent region having a high Reynolds number. By referring to the previous study, the APR+ design is linearly reduced to 1/5 ratio with a 1/2 of the velocity scale, which yields a 1/39.7 of Reynolds number scaling ratio. In the present study, the design feature of the facilities, named 'ACOP', in order to investigate flow and pressure distribution are described

  3. Construction of the neutron beam facility at Australia's OPAL research reactor

    International Nuclear Information System (INIS)

    Kennedy, Shane J.

    2006-01-01

    Australia's new research reactor, OPAL, has been designed principally for neutron beam science and radioisotope production. It has a capacity for 18 neutron beam instruments, located at the reactor face and in a neutron guide hall. The neutron beam facility features a 20 l liquid deuterium cold neutron source and cold and thermal supermirror neutron guides. Nine neutron beam instruments are under development, of which seven are scheduled for completion in early 2007. The project is approaching the hot-commissioning stage, when criticality will be demonstrated. Installation of the neutron beam transport system and neutron beam instruments in the neutron guide hall and at the reactor face is underway, and the path to completion of this project is relatively clear. This paper will outline the key features of the OPAL reactor, and will describe the neutron beam facility in particular. The status of the construction and a forecast of the program to completion, including commissioning and commencement of routine operation in 2007 will also be discussed

  4. Construction of the Neutron Beam Facility at Australia's OPAL Research Reactor

    International Nuclear Information System (INIS)

    Kennedy, J.S.

    2005-01-01

    Full text: Australia's new research reactor, OPAL, has been designed for high quality neutron beam science and radioisotope production. It has a capacity for eighteen neutron beam instruments to be located at the reactor face and in a neutron guide hall. The new neutron beam facility features a 20 litre liquid deuterium cold neutron source and supermirror neutron reflecting guides for intense cold and thermal neutron beams. Nine neutron beam instruments are under development, of which seven are scheduled for completion in early 2007. The project is approaching the hot-commissioning stage, where criticality will be demonstrated. Installation of the neutron beam transport system and neutron beam instruments in the neutron guide hall and at the reactor face is underway, and the path to completion of this project is relatively clear. The lecture will outline Australia's aspirations for neutron science at the OPAL reactor, and describe the neutron beam facility under construction. The status of this project and a forecast of the program to completion, including commissioning and commencement of routine operation in 2007 will also be discussed. This project is the culmination of almost a decade of effort. We now eagerly anticipate catapulting Australia's neutron beam science capability to meet the best in the world today. (author)

  5. Handling of multiassembly sealed baskets between reactor storage and a remote handling facility

    International Nuclear Information System (INIS)

    Massey, J.V.; Kessler, J.H.; McSherry, A.J.

    1989-06-01

    The storage of multiple fuel assemblies in sealed (welded) dry storage baskets is gaining increasing use to augment at-reactor fuel storage capacity. Since this increasing use will place a significant number of such baskets on reactor sites, some initial downstream planning for their future handling scenarios for retrieving multi-assembly sealed baskets (MSBs) from onsite storage and transferring and shipping the fuel (and/or the baskets) to a federally operated remote handling facility (RHF). Numerous options or at-reactor and away-from-reactor handling were investigated. Materials handling flowsheets were developed along with conceptual designs for the equipment and tools required to handle and open the MSBs. The handling options were evaluated and compared to a reference case, fuel handling sequence (i.e., fuel assemblies are taken from the fuel pool, shipped to a receiving and handling facility and placed into interim storage). The main parameters analyzed are throughout, radiation dose burden and cost. In addition to evaluating the handling of MSBs, this work also evaluated handling consolidated fuel canisters (CFCs). In summary, the handling of MSBs and CFCs in the store, ship and bury fuel cycle was found to be feasible and, under some conditions, to offer significant benefits in terms of throughput, cost and safety. 14 refs., 20 figs., 24 tabs

  6. Classification of Reactor Facility Operational State Using SPRT Methods with Radiation Sensor Networks

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez Aviles, Camila A. [ORNL; Rao, Nageswara S. [ORNL

    2018-01-01

    We consider the problem of inferring the operational state of a reactor facility by using measurements from a radiation sensor network, which is deployed around the facility’s ventilation stack. The radiation emissions from the stack decay with distance, and the corresponding measurements are inherently random with parameters determined by radiation intensity levels at the sensor locations. We fuse measurements from network sensors to estimate the intensity at the stack, and use this estimate in a one-sided Sequential Probability Ratio Test (SPRT) to infer the on/off state of the reactor facility. We demonstrate the superior performance of this method over conventional majority vote fusers and individual sensors using (i) test measurements from a network of NaI sensors, and (ii) emulated measurements using radioactive effluents collected at a reactor facility stack. We analytically quantify the performance improvements of individual sensors and their networks with adaptive thresholds over those with fixed ones, by using the packing number of the radiation intensity space.

  7. Present state of inspection robot technology in nuclear power facilities. Case of fast breeder reactors

    International Nuclear Information System (INIS)

    Ara, Kuniaki

    1995-01-01

    In the maintenance works in nuclear power facilities such as checkup, inspection and repair, for the main purpose of radiation protection, remote operation technology was introduced since relatively early stage, and at present, the robots that carry out the inspection works for confirming the soundness of main equipment have been developed and put to practical use. At the time of introducing these technologies, in addition to the research and development of robots proper, the coordination with the design of plant machinery and equipment facilities as the premise of introducing robots is an important requirement. In this report, the present state of the development of remote inspection technology for fast breeder reactors is introduced, and the matters to which attention is paid in the plant design for introducing robots are explained. First, fast breeder reactors are described. The needs of robotizing and adopting remote operation in nuclear power facilities are explained, using the examples of the inspection system for a reactor vessel and the inspection system for steam generator heat transfer tubes. (K.I.)

  8. Use of the TRIGA reactor for the study of neutron pulses effect on semiconductors

    International Nuclear Information System (INIS)

    Di Giorgio, A.; Gallo, G.

    1972-01-01

    For the analysis of the behavior and the response of Si crystal solid state semiconductor detectors to n-radiations and/or gamma pulsating fluxes, experiments are in progress for a long time, using TRIGA of the LENA Reactor. At first, it had been taken into consideration the possibility to use particle accelerators with which, taking advantage of several types of reactions, neutrons fluxes can be obtained with temporal distribution of remarkable interest in the field of the measures of resolution of the sensitive elements. But the possibility of using the accelerators in the field of n-fluxes radiometry appears to be limited as the flux modulators, the Klystron-Reflex ones, do not allow the operation in stable regime with frequencies of order lower than a 10 Hz limit. That causes a remarkable accumulation of heat in the detectors, at the higher radiation fluxes, than does not allow to determine the response range in a linear operation regime; moreover, it turns out to be difficult to obtain a wide energetic spectrum n-flux. The TRIGA reactor, on the contrary, is able to operate, as acknowledged, both at steady state, with variable powers up to 250 KW, and at pulse mode with variable peak powers up to 250 MW and with impulse width around 30 msec, and it's able to supply: elevated wide energetic spectrum n-fluxes; gamma mounts/impulse up to the values of the Mrad; reproducible single impulses; proportionality between n-fluxes and peak power; possibility to control the sensitivity of the detectors in thermal, epithermal and fast groups of the n-flux. Such performances assure an wide range of experimental research, particularly concerning the spectral analysis of the output signals of the detectors, possibly used as transducers in systems measuring n- and/or gamma-flux

  9. 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

  10. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Design of inductively driven long pulse tokamak reactors: IDLT

    International Nuclear Information System (INIS)

    Ogawa, Yuichi

    1998-01-01

    Based on scientific data based adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R and D program, the scientific feasibility of inductively-driven tokamak fusion reactors is studied. A low wall-loading DEMO fusion reactor is designed, which utilizes an austenitic stainless steel in conjunction with significant data bases and operating experiences, since we have given high priority to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the DEMO reactor with the relatively large volume (i.e., major radius of 10 m) is employed, plasma ignition is achievable with a low fusion power of 0.8 GW, and an operation period of 4 - 5 hours is available only with inductive current drive. Disadvantages of pulsed operation in commercial fusion reactors include fatigue in structural materials and the necessity of an energy storage system to compensate the electric power during the dwell time. To overcome these disadvantages, a pulse length is prolonged up to about 10 hours, resulting in the remarkable reduction of the total cycle number to 10 4 during the life of the fusion plant. (author)

  11. Applications: REP-rate pulse power technology

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    Research on the following topics is discussed: (1) REP-rate pulse power technology, (2) RTF-I, 300-J, 100-pps test facility experiments, (3) transformer development, (4) reactor system studies, (5) general conceptual design, (6) economic considerations, (7) specific reactor designs, (8) low-current density diode physics studies, and (9) plasma injected, microsecond, E-beam diodes

  12. The WNR facility - a pulsed spallation neutron source at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

    Russell, G.J.; Lisowski, P.W.; King, N.S.P.

    1978-01-01

    The Weapons Neutron Research facility (WNR) at the Los Alamos Scientific Laboratory is the first operating example of a new class of pulsed neutron sources using the X(p,n)Y spallation reaction. At present, up to 10 microamperes of 800-MeV protons from the Clinton P. Anderson Meson Physics Facility (LAMPF) linear accelerator bombard a Ta target to produce an intense white-neutron spectrum from about 800 MeV to 100 keV. The Ta target can be coupled with CH 2 and H 2 O moderators to produce neutrons of lower energy. The time structure of the WNR proton beam may be varied to optimize neutron time-of-flight (TOF) measurements covering the energy range from several hundred MeV to a few meV. The neutronics of the WNR target and target/moderator configurations have been calculated from 800 MeV to 0.5 eV. About 11 neutrons per proton are predicted for the existing Ta target. Some initial neutron TOF data are presented and compared with calculations

  13. A 10-TW Pulsed Facility "PIRIT" for Investigation of Short-Wave Radiation Generation.

    Science.gov (United States)

    Popkov, N F; Ryaslov, E A; Kargin, V I; Pikar', A S; Kotel'nikov, D V; Melkozerov, A V

    1995-01-01

    The results of experiments with a plasma x-ray source in the PIRIT-2000 facility are presented in this paper. The facility is designed with module capacitive energy storage energizing vacuum inductive storage. The formation of a rapidly growing current pulse as well as its commutation on a load was carried out by a plasma opening switch. A vacuum diode as well as various types of plasma loads can be used for the generation of a high-power x-ray flux. The storage energy of a 54-module capacitive storage is up to 2 MJ, its inductance is 15 nH, and its output voltage is 500 kV. The peak current in the plasma load constituted 4 MA with a 150-ns rise time. The maximum integral energy output of x radiation measured by an open thermocouple calorimeter was as high as 100 kJ, while the primary storage energy was 1 MJ. The plasma load usage at a current of 4 MA ensured a 100-kJ generation in x-ray radiation and the density of the radiation flux at a distance of 1 m from the source was as much as 0.8 J/cm2, while near the source it was 10 J/cm2.

  14. Analytical evaluation of neutron diffusion equation for the geometry of very intense continuous high flux pulsed reactor

    International Nuclear Information System (INIS)

    Narain, Rajendra

    1995-01-01

    Using the concept of Very Intense Continuous High Flux Pulsed Reactor to obtain a rotating high flux pulse in an annular core an analytical treatment for the quasi-static solution with a moving reflector is presented. Under quasi-static situation, time averaged values for important parameters like multiplication factor, flux, leakage do not change with time. As a result the instantaneous solution can be considered to be separable in time and space after correcting for the coordinates for the motion of the pulser. The space behaviour of the pulser is considered as exp(-αx 2 ). Movement of delayed neutron precursors is also taken into account. (author). 4 refs

  15. The Budapest research reactor as an advanced research facility for the early 21st century

    International Nuclear Information System (INIS)

    Vidovszky, I.

    2001-01-01

    The Budapest Research Reactor, Hungary's first nuclear facility was originally put into operation in 1959. The reactor serves for: basic and applied research, technological and commercial applications, education and training. The main goal of the reactor is to serve neutron research. This unique research possibility is used by a broad user community of Europe. Eight instruments for neutron scattering, radiography and activation analyses are already used, others (e.g. time of flight spectrometer, neutron reflectometer) are being installed. The majority of these instruments will get a much improved utilization when the cold neutron source is put into operation. In 1999 the Budapest Research Reactor was operated for 3129 full power hours in 14 periods. The normal operation period took 234 hours (starting Monday noon and finishing Thursday morning). The entire production for the year 1999 was 1302 MW days. This is a slightly reduced value, due to the installation of the cold neutron source. For the year 2000 a somewhat longer operation is foreseen (near to 4000 hours), as the cold neutron source will be operational. The operation of the reactor is foreseen at least up to the end of the first decade of the 21 st century. (author)

  16. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles

  17. Proposed power upgrade of the hot fuel examination facility's neutron radiography reactor

    International Nuclear Information System (INIS)

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1984-01-01

    The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both non-destructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the non-destructive examination techniques utilized at HFEF is neutron radiography. When the NRAD facility was designed and constructed, an operating power level of 250 kw was considered to be adequate for obtaining radiographs of the type of specimens envisaged at that time. Since that time, several things have occurred that have tended to increase radiography exposure times to as much as 90 minutes each. In order to decrease exposure times, the reactor power level is to be increased from 250 kW to 1 MW. This increase in power will necessitate several engineering and design changes. The proposed upgrade of the NRAD facility will increase the neutron flux available in the beam tubes appreciably. The increased flux will enable NRAD to continue to meet its operational commitments in a timely manner and to develop state-of-the-art techniques in the future as it has in the past

  18. A pulsed neutron monochromatic beam at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Adib, M.; Abdel-Kawy, A.; Eid, Y.; Maayouf, R.M.A.

    1985-01-01

    A pulsed neutron monochromatic beam, at the ET-RR-1 reactor, is produced by two 32 cm diameter rotors suspended in magnetic fields, whose centres are 126 cm apart rotating at speeds up to 16,000 rev/min. Each of the rotors has two slots, which are of constant cross-section in area - 7x10mm 2 , and are curved so that they have a maximum transmission for neutrons whose speed is 8.2 times that of the rotor tip. The jitter of the phase between the rotors at different rotation rates is found not to exceed +-1 μs. It has been found that both the observed time distribution and the TOF distribution of the neutrons at different rotation rates are in good agreement with the calculated ones. The observed intensity of the monochromatic neutrons of wavelength 2.74+-0.09 A, obtained by the rotors rotating at a speed of 10,500 rev/min with 864+-1 μs difference in phase between them, is 66.8 n/s. This value is found to be less than the predicted one by a factor of 5.5. (author)

  19. 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.)

  20. Megajoule-class single-pulse KrF laser test facility as a logical step toward inertial fusion commercialization

    International Nuclear Information System (INIS)

    Harris, D.B.; Pendergrass, J.H.

    1985-01-01

    The cost and efficiency of megajoule-class KrF laser single pulse test facilities have been examined. A baseline design is described which illuminates targets with 5 MJ with shaped 10-ns pulses. The system uses 24 main amplifiers and operates with an optics operating fluence of 4.0 J/cm 2 . This system has 9.0% efficiency and costs $200/joule. Tradeoff studies indicate that large amplifier modules and high fluences lead to the lowest laser system costs, but that only a 20% cost savings can be realized by going to amplifier modules larger than 200 kJ and/or fluences greater than 4 J/cm 2 . The role of the megajoule-class single-pulse test facility towards inertial fusion commercialization will also be discussed

  1. Post 9-11 Security Issues for Non-Power Reactor Facilities

    International Nuclear Information System (INIS)

    Zaffuts, P. J.

    2003-01-01

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so

  2. Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

    International Nuclear Information System (INIS)

    Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

    1977-01-01

    An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

  3. Advantages of co-located spent fuel reprocessing, repository and underground reactor facilities

    International Nuclear Information System (INIS)

    Mahar, James M.; Kunze, Jay F.; Wes Myers, Carl; Loveland, Ryan

    2007-01-01

    The purpose of this work is to extend the discussion of potential advantages of the underground nuclear park (UNP) concept by making specific concept design and cost estimate comparisons for both present Generation III types of reactors and for some of the modular Gen IV or the GNEP modular concept. For the present Gen III types, we propose co-locating reprocessing and (re)fabrication facilities along with disposal facilities in the underground park. The goal is to determine the site costs and facility construction costs of such a complex which incorporates the advantages of a closed fuel cycle, nuclear waste repository, and ultimate decommissioning activities all within the UNP. Modular power generation units are also well-suited for placement underground and have the added advantage of construction using current and future tunnel boring machine technology. (authors)

  4. Post 9-11 Security Issues for Non-Power Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Zaffuts, P. J.

    2003-02-25

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so.

  5. Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor

    CERN Document Server

    Int. At. Energy Agency, Wien

    2002-01-01

    This document describes the specific dispositions relative to the nuclear reactor domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor and of the retraining, the Passerelle formation, are presented. (A.L.B.)

  6. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  7. Conceptual design report for the away from reactor spent fuel storage facility, Savannah River Plant

    International Nuclear Information System (INIS)

    1978-12-01

    The Department of Energy (DOE) requested that Du Pont prepare a conceptual design and appraisal of cost for Federal budget planning for an away from reactor spent fuel storage facility that could be ready to store fuel by December 1982. This report describes the basis of the appraisal of cost in the amount of $270,000,000 for all facilities. The proposed action is to provide a facility at the Savannah River Plant. The facility will have an initial storage capacity of 5000 metric tons of spent fuel and will be capable of receiving 1000 metric tons per year. The spent fuel will be stored in water-filled concrete basins that are lined with stainless steel. The modular construction of the facility will allow future expansion of the storage basins and auxiliary services in a cost-effective manner. The facility will be designed to receive, handle, decontaminate and reship spent fuel casks; to remove irradiated fuel from casks; to place the fuel in a storage basin; and to cool and control the quality of the water. The facility will also be designed to remove spent fuel from storage basins, load the spent fuel into shipping casks, decontaminated loaded casks and ship spent fuel. The facility requires a license by the Nuclear Regulatory Commission (NRC). Features of the design, construction and operations that may affect the health and safety of the workforce and the public will conform with NRC requirements. The facility would be ready to store fuel by January 1983, based on normal Du Pont design and construction practices for DOE. The schedule does not include the effect of licensing by the NRC. To maintain this option, preparation of the documents and investigation of a site at the Savannah River Plant, as required for licensing, were started in FY '78

  8. Perspectives and possibilities for solid state physcis investigations at thhe pulsed reactor IBR-2 of the JINR Dubna

    International Nuclear Information System (INIS)

    Feldmann, K.; Frauenheim, T.; Lauckner, J.; Weniger, J.; Muehle, E.

    1982-02-01

    Three time-of-flight spectrometers (spectrometer of high resolution NSWR, spectrometer of inverted geometry KDSOG and spectrometer of polarized neutrons SPN-1) are presented, which will be working at the pulsed IBR-2 reactor of the JINR Dubna. Considering the parameters and the special methods of measurement of these spectrometers, the possibilities of their applications for the investigations of structural, magnetic and electronic properties of solids by means of elastic, inelastic and quasielastic neutron scattering are discussed. (author)

  9. Basic features of electromagnetic pulse generated in a laser-target chamber at 3-TW laser facility PALS

    International Nuclear Information System (INIS)

    De Marco, M; Pfeifer, M; Krousky, E; Krasa, J; Cikhardt, J; Klir, D; Nassisi, V

    2014-01-01

    We describe the radiofrequency emission taking place when 300 ps laser pulses irradiate various solid targets with an intensity of 10 16 W/cm 2 . The emission of intense electromagnetic pulses was observed outside the laser target chamber by two loop antennas up to 1 GHz. Electromagnetic pulses can be 800 MHz transients, which decay from a peak electromagnetic field of E 0 ≊ 7 kV/m and H 0 ≊ 15 A/m. The occurrence of these electromagnetic pulses is associated with generation of hard x-rays with photon energies extending beyond 1 MeV. This contribution reports the first observation of this effect at the PALS facility.

  10. Russian Federation: Passive Safety Components for Lead-Cooled Reactor Facilities

    International Nuclear Information System (INIS)

    Sarkulov, M.K.

    2015-01-01

    There is a specific range of engineered features used traditionally in nuclear technology. As a rule, main reactivity control systems use conventional active actuators with solid-body control members and/or liquid systems with active injection of liquid absorber. Other operation principles are normally chosen for additional systems. Currently, the traditional approach to improving the reliability of a reactor facility suggests an increase in the number of safety components and systems which provide for mutual assurance or assist each other. There is a great variety of additional reactivity control members designed for the reactor facility control and shutdown, including hydrodynamic members in the form of rods (acting from the coolant flow); floating-type members (absorbers and displacers); storage-type and liquid members (used in separate channels); bulk members (pebble absorber); gas-based members (with a gas absorber); shape-memory members and others. Hydrodynamic systems were introduced at Beloyarsk NPP Units 1 and 2 and proposed for use in other facility designs, Gases and bulk materials have not been commonly accepted: the former because of the high cost of high-efficiency gaseous absorbers, and the latter because of the complecated monitoring of the bulk material position. It is rather difficult and not always necessary to use the same engineering approaches in new lead-cooled reactor facilities as in traditional ones. Similarly to the development of traditional safety systems, passive safety components (devices) shall be designed according to the essential requirements of the nuclear regulations of the Russian Federation

  11. Decomposition of methane to hydrogen using nanosecond pulsed plasma reactor with different active volumes, voltages and frequencies

    International Nuclear Information System (INIS)

    Khalifeh, Omid; Mosallanejad, Amin; Taghvaei, Hamed; Rahimpour, Mohammad Reza; Shariati, Alireza

    2016-01-01

    Highlights: • CH 4 conversion into H 2 is investigated in a nanosecond pulsed DBD reactor. • The absence of CO and CO 2 in the product gas is highly favorable. • Effects of external electrode length, applied voltage and frequency are examined. • The maximum efficiency of 7.23% is achieved at the electrode length of 15 cm. • The maximum CH 4 conversion of 87.2% is obtained at discharge power 268.92 W. - Abstract: In this paper, the methane conversion into hydrogen is investigated experimentally in a nanosecond pulsed DBD reactor. In order to achieve pure hydrogen production with minimum power consumption, effects of some operating parameters including external electrode length, applied voltage and pulse repetition frequency have been evaluated. Results show that although higher CH 4 conversion and H 2 concentration can be obtained at longer electrode lengths, higher applied voltages and pulse repetition frequencies, these parameters should be optimized for efficient hydrogen production. Actually, the maximum CH 4 conversion of 87.2% and maximum hydrogen percentage of 80% are obtained at the external electrode length, discharge power, voltage and frequency of 15 cm, 268.92 W, 12 kV and 10 kHz, respectively. However, the maximum efficiency of 7.23% is achieved at the external electrode length of 15 cm, applied voltage of 6 kV, pulse repetition frequency of 0.9 kHz and discharge power of 4 W. Furthermore, at this condition, due to low temperature of discharge zone very little amount of solid carbon was observed on the inner electrode surface of the reactor.

  12. Development of protection system for power supply facilities in JT-60U P-NBI for long pulse operation

    International Nuclear Information System (INIS)

    Ohshima, Katsumi; Okano, Fuminori; Honda, Atsushi; Shinozaki, Shin-ichi; Usui, Katsutomi; Noto, Katsuya; Kawai, Mikito; Ikeda, Yoshitaka

    2007-06-01

    In the positive ion based NBI (P-NBI) system, we have developed a protection system to protect the power supply facilities from over load during long pulse operation. The protection system monitors the voltage (V) and current (I) in the power supply facilities, and calculates the parameters of V2t and I2t in real-time, where T is the pulse duration. It turns off the power supply facilities when V2t and I2t are beyond the critical values. After two development stages, we have completed the protection system using a package typed PLC (Programmable Logic Controller) which has a high expandability of multi-unit operation. Moreover, we have constructed a user-friendly system by using a SCADA (Supervisory Control and Data Acquisition) system. (author)

  13. Modular head assembly and method of retrofitting existing nuclear reactor facilities

    International Nuclear Information System (INIS)

    Malandra, L.J.; Ledue, R.J.; Hankinson, M.F.; Kowalski, E.F.

    1987-01-01

    A method is described of retrofitting existing nuclear reactor facilities so as to form a modular closure head assembly for a nuclear reactor pressure vessel, where the existing nuclear reactor facilities comprise control rod drive mechanism cooling systems which include vertically extending elbow air ducts inter-connecting vertically spaced upper and lower air manifolds. The elbow air ducts extend radially beyond the peripheral envelope of the closure head, comprising the steps of: removing the upper air manifold; removing the vertically extending elbow air ducts; capping the air ports of the lower air manifold which ports were previously fluidically connecting the lower air manifold to the vertically extending elbow air ducts; disposing vertically upwardly extending air exhaust ducts above the lower air manifold in such an manner that the air exhaust ducts are disposed within the peripheral envelope of the closure head; fluidically connecting exhaust fans to the upper regions of the air exhaust ducts; fluidically connecting the lower regions of the air exhaust ducts the lower air manifold; permanently securing lift rods to the closure head at positions disposed radially outwardly of the lower air manifold; attaching a seismic support platform to the lift rods; proving fluidic passage of the vertically extending air exhaust ducts through the seismic support platform; attaching a missile shield plate to the lift rods; and proving fluidic passage of the vertically extending air exhaust ducts through the missile shield plate

  14. Applied research and service activities at the University of Missouri Research Reactor Facility (MURR)

    International Nuclear Information System (INIS)

    Alger, D.M.

    1987-01-01

    The University Of Missouri operates MURR to provide an intense source of neutron and gamma radiation for research and applications by experimenters from its four campuses and by experimenters from other universities, government and industry. The 10 MW reactor, which has been operating an average of 155 hours per week for the past eight years, produces thermal neutron fluxes up to 6-7x10 14 n/cm 2 -s in the central flux trap and beamport source fluxes of up to 1.2x10 14 n/cm 2 -s. The mission of the reactor facility, to promote research, education and service, is the same as the overall mission of the university and therefore, applied research and service supported by industrial firms have been welcomed. The university recognized after a few years of reactor operation that in order to build utilization, it would be necessary to develop in-house research programs including people, equipment and activity so that potential users could more easily and quickly obtain the results needed. Nine research areas have been developed to create a broadly based program to support the level of activity needed to justify the cost of operating the facility. Applied research and service generate financial support for about one-half of the annual budget. The applied and service programs provide strong motivation for university/industry association in addition to the income generated. (author)

  15. Laser stand for irradiation of targets by laser pulses from the Iskra-5 facility at a repetition rate of 100 MHz

    International Nuclear Information System (INIS)

    Annenkov, V I; Garanin, Sergey G; Eroshenko, V A; Zhidkov, N V; Zubkov, A V; Kalipanov, S V; Kalmykov, N A; Kovalenko, V P; Krotov, V A; Lapin, S G; Martynenko, S P; Pankratov, V I; Faizullin, V S; Khrustalev, V A; Khudikov, N M; Chebotar, V S

    2009-01-01

    A train of a few tens of high-power subnanosecond laser pulses with a repetition period of 10 ns is generated in the Iskra-5 facility. The laser pulse train has an energy of up to 300 J and contains up to 40 pulses (by the 0.15 intensity level), the single pulse duration in the train being ∼0.5 ns. The results of experiments on conversion of a train of laser pulses to a train of X-ray pulses are presented. Upon irradiation of a tungsten target, a train of X-ray pulses is generated with the shape of an envelope in the spectral band from 0.18 to 0.28 keV similar to that of the envelope of the laser pulse train. The duration of a single X-ray pulse in the train is equal to that of a single laser pulse. (lasers)

  16. Pulse-density modulation control of chemical oscillation far from equilibrium in a droplet open-reactor system.

    Science.gov (United States)

    Sugiura, Haruka; Ito, Manami; Okuaki, Tomoya; Mori, Yoshihito; Kitahata, Hiroyuki; Takinoue, Masahiro

    2016-01-20

    The design, construction and control of artificial self-organized systems modelled on dynamical behaviours of living systems are important issues in biologically inspired engineering. Such systems are usually based on complex reaction dynamics far from equilibrium; therefore, the control of non-equilibrium conditions is required. Here we report a droplet open-reactor system, based on droplet fusion and fission, that achieves dynamical control over chemical fluxes into/out of the reactor for chemical reactions far from equilibrium. We mathematically reveal that the control mechanism is formulated as pulse-density modulation control of the fusion-fission timing. We produce the droplet open-reactor system using microfluidic technologies and then perform external control and autonomous feedback control over autocatalytic chemical oscillation reactions far from equilibrium. We believe that this system will be valuable for the dynamical control over self-organized phenomena far from equilibrium in chemical and biomedical studies.

  17. Radiological controls and worker and public health and safety: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Tew, J.L.; Miles, M.E.; Knuth, D.; Boyd, R.

    1981-02-01

    DOE has formed a Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee to assess the implications of the Report of the President's Commission on the Accident at Three Mile Island that are applicable to DOE's nuclear reactor operations. Thirteen DOE nuclear reactors were reviewed by the Committee. This report was prepared to provide a measure of how the radiological control and environmental practices at the 13 individual DOE reactor facilities measure up to (1) the recommendations contained in the Report of the President's Commission on the Accident at Three Mile Island, (2) the requirements and guidelines contained, and (3) the requirements of the applicable Title and Part of the Code of Federal Regulations

  18. Comparative study on various attenuation theories usually used for aseismatic design of reactor facilities

    International Nuclear Information System (INIS)

    Muto, Kiyoshi; Kobayashi, Toshio.

    1977-01-01

    In the vibration analysis for the aseismatic design of reactor facilities, various attenuation theories have been used case by case. In this study, three attenuation theories of internal viscosity type, strain energy type and kinetic energy type were applied to the vibration analysis for a BWR type reactor building, and the respective results were compared. The three theories are explained briefly. Different damping factors were taken for the RC building, steel internal equipments and ground. The reactor building for a BWR of Mark 1 type was studied, and the mathematic vibration model involving those for the building, shield wall, primary containment vessel, gamma shield wall, reactor pressure vessel and foundation is shown. Natural frequency analysis, frequency response analysis and earthquake response analysis were carried out. Regarding the response of the building, the internal viscosity type and the strain energy type gave almost same values, but the kinetic energy type gave the value 1.2 times as large as the former values. In the response of the building, the first mode element is predominant, and the damping factors to the first mode caused these results. Regarding the maximum response acceleration and floor response spectra of internal equipments, the strain energy type and the kinetic energy type gave similar results, and the internal viscosity type gave smaller values. (Kako, I.)

  19. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.

    1986-01-01

    A new conceptual design of a fusion reactor blanket simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBR), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation methods because design deficiencies that could affect results are insignificant. Inelastic scattering cross sections are identified as a major source of these discrepancies. The conceptual design of this FBBR analog, the fusion reactor blanket facility (FRBF), is presented. Essential features are a cylindrical geometry and a distributed, cosine-shaped line source of 14-MeV neutrons. This source can be created by sweeping a deuteron beam over an elongated titanium-tritide target. To demonstrate that the design of the FRBF will not contribute significant deviations in experimental results, neutronics analyses were performed: results of comparisons of 2-dimensional to 1-dimensional predictions are reported for two blanket compositions. Expected deviations from 1-D predictions which are due to source anisotropy and blanket asymmetry are minimal. Thus, design of the FRBF allows simple and straightforward interpretation of the experimental results, without a need for coarse 3-D calculations

  20. Planning and management for the decommissioning of research reactors and other small nuclear facilities

    International Nuclear Information System (INIS)

    1993-01-01

    Many research reactors and other small nuclear facilities throughout the world date from the original nuclear research programmes in the Member States. Consequently, a large number of these plants have either been retired from service or will soon reach the end of their useful lives and are likely to become significant decommissioning tasks for those Members States. In recognition of this situation and in response to considerable interest shown by Member States, the IAEA has produced this document on planning and management for the decommissioning of research reactors and other small nuclear facilities. While not directed specifically at large nuclear installations, it is likely that much of the information presented will also be of interest to those involved in the decommissioning of such facilities. Current views, information and experience on the planning and management of decommissioning projects in Member States were collected and assessed during a Technical Committee Meeting held by the IAEA in Vienna from 29 July to 2 August 1991. It was attended by 22 participants from 14 Member States and one international organization. 28 refs, 2 figs, 3 tabs