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Sample records for solid core reactors

  1. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  2. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

  3. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  4. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  5. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

  6. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  7. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Diaz, N.J.; Dugan, E.T.

    1983-01-01

    A heterogeneous gas core nuclear reactor is disclosed comprising a core barrel provided interiorly with an array of moderator-containing tubes and being otherwise filled with a fissile and/or fertile gaseous fuel medium. The fuel medium may be flowed through the chamber and through an external circuit in which heat is extracted. The moderator may be a fluid which is flowed through the tubes and through an external circuit in which heat is extracted. The moderator may be a solid which may be cooled by a fluid flowing within the tubes and through an external heat extraction circuit. The core barrel is surrounded by moderator/coolant material. Fissionable blanket material may be disposed inwardly or outwardly of the core barrel

  8. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  9. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

    A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

  10. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  11. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Han, K.I.

    1977-01-01

    Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

  12. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  13. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Reactor core cooling device

    International Nuclear Information System (INIS)

    Kobayashi, Masahiro.

    1986-01-01

    Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

  15. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

    Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

  16. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

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

  17. Reactor core control device

    International Nuclear Information System (INIS)

    Sano, Hiroki

    1998-01-01

    The present invention provides a reactor core control device, in which switching from a manual operation to an automatic operation, and the control for the parameter of an automatic operation device are facilitated. Namely, the hysteresis of the control for the operation parameter by an manual operation input means is stored. The hysteresis of the control for the operation parameter is collected. The state of the reactor core simulated by an operation control to which the collected operation parameters are manually inputted is determined as an input of the reactor core state to the automatic input means. The record of operation upon manual operation is stored as a hysteresis of control for the operation parameter, but the hysteresis information is not only the result of manual operation of the operation parameter. This is results of operation conducted by a skilled operator who judge the state of the reactor core to be optimum. Accordingly, it involves information relevant to the reactor core state. Then, it is considered that the optimum automatic operation is not deviated greatly from the manual operation. (I.S.)

  18. Plasma core reactor applications

    International Nuclear Information System (INIS)

    Latham, T.S.; Rodgers, R.J.

    1976-01-01

    Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

  19. Nuclear reactor core assembly

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1978-01-01

    The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

  20. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  1. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

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

  2. Initial charge reactor core

    International Nuclear Information System (INIS)

    Kiyono, Takeshi

    1984-01-01

    Purpose: To effectivity burn fuels and improve the economical performance in an inital charge reactor core of BWR type reactors or the likes. Constitution: In a reactor core constituted with a plurality of fuel assemblies which are to be partially replaced upon fuel replacement, the density of the fissionable materials and the moderator - fuel ratio of a fuel assembly is set corresponding to the period till that fuel assembly is replaced, in which the density of the nuclear fissionable materials is lowered and the moderator - fuel ratio is increased for the fuel assembly with a shorter period from the fueling to the fuel exchange and, while on the other hand, the density of the fissionable materials is increased and the moderator - fuel ratio is decreased for the fuel assembly with a longer period from the fueling to the replacement. Accordingly, since the moderator - fuel ratio is increased for the fuel assembly to be replaced in a shorter period, the neutrons moderating effect is increased to increase the reactivity. (Horiuchi, T.)

  3. Nuclear reactor core

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo; Ishibashi, Yoko; Mochida, Takaaki; Haikawa, Katsumasa; Yamanaka, Akihiro.

    1995-01-01

    A reactor core is radially divided into an inner region, an outer region and an outermost region. As a fuel, three kinds of fuels, namely, a high enrichment degree fuel at 3.4%, a middle enrichment degree fuel at 2.3% and a low enrichment degree at 1.1% of a fuel average enrichment degree of fission product are used. Each of the fuels is bisected to upper and lower portions at an axial center thereof. The difference of average enrichment degrees between upper and lower portions is 0.1% for the high enrichment degree fuel, 0.3% for the middle enrichment degree fuel and 0.2% for the low enrichment degree fuel. In addition, the composition of fuels in each of radial regions comprises 100% of the low enrichment degree fuels in the outermost region, 91% of the higher enrichment degree fuels and 9% of the middle enrichment degree fuels in the outer region, and 34% of the high enrichment degree fuels and 30% of the middle enrichment degree fuels in the inner region. With such a constitution, fuel economy can be improved while maintaining the thermal margin in an initially loaded reactor core of a BWR type reactor. (I.N.)

  4. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

    A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

  5. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  6. RB reactor benchmark cores

    International Nuclear Information System (INIS)

    Pesic, M.

    1998-01-01

    A selected set of the RB reactor benchmark cores is presented in this paper. The first results of validation of the well-known Monte Carlo MCNP TM code and adjoining neutron cross section libraries are given. They confirm the idea for the proposal of the new U-D 2 O criticality benchmark system and support the intention to include this system in the next edition of the recent OECD/NEA Project: International Handbook of Evaluated Criticality Safety Experiment, in near future. (author)

  7. 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.)

  8. Nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F

    1974-07-11

    The core of the fast neutron reactor consisting, among other components, of fuel elements enriched in plutonium is divided into modules. Each module contains a bundle of four or six elongated components (fuel elements and control rods). In the arrangement with four components, one is kept rigid while the other three are elastically yielding inclined towards the center and lean against the rigid component. In the modules with six pieces, each component is elastically yielding inclined towards a central cavity. In this way, they form a circular arc. A control rod may be placed in the cavity. In order to counteract a relative lateral movement, the outer surfaces of the components which have hexagonal cross-sections have interlocking bearing cushions. The bearing cushions consist of keyway-type ribs or grooves with the wedges or ribs gripping in the grooves of the neighbouring components. In addition, the ribs have oblique entering surfaces.

  9. Reactor core performance calculating device

    International Nuclear Information System (INIS)

    Tominaga, Kenji; Bando, Masaru; Sano, Hiroki; Maruyama, Hiromi.

    1995-01-01

    The device of the present invention can calculate a power distribution efficiently at high speed by a plurality of calculation means while taking an amount of the reactor state into consideration. Namely, an input device takes data from a measuring device for the amount of the reactor core state such as a large number of neutron detectors disposed in the reactor core for monitoring the reactor state during operation. An input data distribution device comprises a state recognition section and a data distribution section. The state recognition section recognizes the kind and amount of the inputted data and information of the calculation means. The data distribution section analyzes the characteristic of the inputted data, divides them into a several groups, allocates them to each of the calculation means for the purpose of calculating the reactor core performance efficiently at high speed based on the information from the state recognition section. A plurality of the calculation means calculate power distribution of each of regions based on the allocated inputted data, to determine the power distribution of the entire reactor core. As a result, the reactor core can be evaluated at high accuracy and at high speed irrespective of the whole reactor core or partial region. (I.S.)

  10. Nuclear reactor with several cores

    International Nuclear Information System (INIS)

    Swars, H.

    1977-01-01

    Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

  11. Reactor core simulations in Canada

    International Nuclear Information System (INIS)

    Roy, R.; Koclas, J.; Shen, W.; Jenkins, D. A.; Altiparmakov, D.; Rouben, B.

    2004-01-01

    This review will address the current simulation flow-chart currently used for reactor-physics simulations in the Canadian industry. The neutron behaviour in heavy-water moderated power reactors is quite different from that in other power reactors, thus the core physics approximations are somewhat different Some codes used are particular to the context of heavy-water reactors, and the paper focuses on this aspect. The paper also shows simulations involving new design features of the Advanced Candu Reactor TM (ACR TM), and provides insight into future development, expected in the coming years. (authors)

  12. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  13. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  14. Nuclear reactor core stabilizing arrangement

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1976-01-01

    A nuclear reactor core stabilizing arrangement is described wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure. 7 claims, 4 drawing figures

  15. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Yuchi, Yoko; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro; Koyama, Jun-ichi.

    1996-01-01

    In a fuel assembly of a BWR type reactor, a region substantially containing burnable poison is divided into an upper region and a lower region having different average concentrations of burnable poison along a transverse cross section perpendicular to the axial direction. The ratio of burnable poison contents of both regions is determined to not more than 80%, and the average concentration of the burnable poison in the lower region is determined to not less than 9% by weight. An infinite multiplication factor at an initial stage of the burning of the fuel assembly is controlled effectively by the burnable poisons. Namely, the ratio of the axial power can be controlled by the distribution of the enrichment degree of uranium fuels and the distribution of the burnable poison concentration in the axial direction. Since the average enrichment degree of the reactor core has to be increased in order to provide an initially loaded reactor core at high burnup degree. Distortion of the power distribution in the axial direction of the reactor core to which fuel assemblies at high enrichment degree are loaded is flattened to improve thermal margin, to extend continuous operation period and increase a burnup degree upon take-out thereby improving fuel economy without worsening the reactor core characteristics of the initially loaded reactor core. (N.H.)

  16. Reactor core performance estimating device

    International Nuclear Information System (INIS)

    Tanabe, Akira; Yamamoto, Toru; Shinpuku, Kimihiro; Chuzen, Takuji; Nishide, Fusayo.

    1995-01-01

    The present invention can autonomously simplify a neural net model thereby enabling to conveniently estimate various amounts which represents reactor core performances by a simple calculation in a short period of time. Namely, a reactor core performance estimation device comprises a nerve circuit net which divides the reactor core into a large number of spacial regions, and receives various physical amounts for each region as input signals for input nerve cells and outputs estimation values of each amount representing the reactor core performances as output signals of output nerve cells. In this case, the nerve circuit net (1) has a structure of extended multi-layered model having direct coupling from an upper stream layer to each of downstream layers, (2) has a forgetting constant q in a corrected equation for a joined load value ω using an inverse error propagation method, (3) learns various amounts representing reactor core performances determined using the physical models as teacher signals, (4) determines the joined load value ω decreased as '0' when it is to less than a predetermined value upon learning described above, and (5) eliminates elements of the nerve circuit net having all of the joined load value decreased to 0. As a result, the neural net model comprises an autonomously simplifying means. (I.S.)

  17. Nuclear reactor core safety device

    International Nuclear Information System (INIS)

    Colgate, S.A.

    1977-01-01

    The danger of a steam explosion from a nuclear reactor core melt-down can be greatly reduced by adding a gasifying agent to the fuel that releases a large amount of gas at a predetermined pre-melt-down temperature that ruptures the bottom end of the fuel rod and blows the finely divided fuel into a residual coolant bath at the bottom of the reactor. This residual bath should be equipped with a secondary cooling loop

  18. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

    1990-01-01

    The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

  19. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

    In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

  20. WWER-440 type reactor core

    International Nuclear Information System (INIS)

    Mizov, J.; Svec, P.; Rajci, T.

    1987-01-01

    Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

  1. One dimensional reactor core model

    International Nuclear Information System (INIS)

    Kostadinov, V.; Stritar, A.; Radovo, M.; Mavko, B.

    1984-01-01

    The one dimensional model of neutron dynamic in reactor core was developed. The core was divided in several axial nodes. The one group neutron diffusion equation for each node is solved. Feedback affects of fuel and water temperatures is calculated. The influence of xenon, boron and control rods is included in cross section calculations for each node. The system of equations is solved implicitly. The model is used in basic principle Training Simulator of NPP Krsko. (author)

  2. Wire core reactor for NTP

    International Nuclear Information System (INIS)

    Harty, R.B.

    1991-01-01

    The development of the wire core system for Nuclear Thermal Propulsion (NTP) that took place from 1963 to 1965 is discussed. A wire core consists of a fuel wire with spacer wires. It's an annular flow core having a central control rod. There are actually four of these, with beryllium solid reflectors on both ends and all the way around. Much of the information on the concept is given in viewgraph form. Viewgraphs are presented on design details of the wire core, the engine design, engine weight vs. thrust, a technique used to fabricate the wire fuel element, and axial temperature distribution

  3. Reactor core monitoring device

    International Nuclear Information System (INIS)

    Ishii, Takanobu; Handa, Hiroaki; Hayashi, Katsumi; Narita, Hitoshi; Shimozaki, Takaaki

    1995-01-01

    The device of the present invention reliably and conveniently detects an event of rapid increase of a coolant void coefficient at a portion of a channel by flow channel clogging event in a PWR-type reactor. Namely, upon flow channel clogging event, the coolant void coefficient is increased, an effective density is lowered, and a coolant shielding effect is lowered. Therefore, fast neutron fluxes at the periphery of a pressure tube are increased. The increase of the fast neutron fluxes is detected by a fast neutron flux detector disposed in a guide tube of an existent neutron flux detector. Based on the result, increase of coolant void coefficient can be detected. When an average void coefficient reaches from 30% to 100%, for example, the fast neutron fluxes are increased by about twice at a neutron permeation distance of coolants of about 10cm, thereby enabling to perform effective detection. (I.S.)

  4. Reactor core operation management system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Tomomi.

    1992-05-28

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.).

  5. Modeling of the reactor core

    International Nuclear Information System (INIS)

    1999-01-01

    In order to improve technical - economical parameters fuel with 2.4% enrichment and burnable absorber is started to be used at Ignalina NPP. Using code QUABOX/CUBBOX the main neutronic - physical characteristics were calculated for selected reactor core conditions

  6. Reactor core operation management system

    International Nuclear Information System (INIS)

    Sato, Tomomi.

    1992-01-01

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.)

  7. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

    Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

  8. Reactor-core-reactivity control device

    International Nuclear Information System (INIS)

    Miura, Teruo; Sakuranaga, Tomonobu.

    1983-01-01

    Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

  9. Fast reactor core monitoring device

    International Nuclear Information System (INIS)

    Sanda, Toshio; Inoue, Kotaro; Azekura, Kazuo.

    1982-01-01

    Purpose: To enable the rapid and accurate on-line identification of the state of a fast reactor core by effectively utilizing the measured data on the temperature and flow rate of the coolant. Constitution: The spacial power distribution and average assembly power are quickly calculated using an approximate calculating method, the measured values and the calculated values of the inlet and outlet temperature difference, flow rate and coolant physical values of an assembly are combined and are individually obtained, the most definite respective values and their errors are obtained by a least square method utilizing a formula of the relation between these values, and the power distribution and the temperature distribution of a reactor core are estimated in this manner. Accordingly, even when the measuring accuracy and the calculating accuracy are equal as in a fast reactor, the power distribution and the temperature distribution can be accurately estimated on-line at a high speed in a nuclear reactor, information required for the operator is provided, and the reactor can thus be safely and efficiently operated. (Yoshihara, H.)

  10. Power reactor core safety research

    International Nuclear Information System (INIS)

    Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

    1981-01-01

    As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

  11. 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.)

  12. 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.)

  13. Nuclear reactor core servicing apparatus

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved core servicing apparatus for a nuclear reactor of the type having a reactor vessel, a vessel head having a head penetration therethrough, a removable plug adapted to fit in the head penetration, and a core of the type having an array of elongated assemblies. The improved core servicing apparatus comprises a plurality of support columns suspended from the removable plug and extending downward toward the nuclear core, rigid support means carried by each of the support columns, and a plurality of servicing means for each of the support columns for servicing a plurality of assemblies. Each of the plurality of servicing means for each of the support columns is fixedly supported in a fixed array from the rigid support means. Means are provided for rotating the rigid support means and servicing means between condensed and expanded positions. When in the condensed position, the rigid support means and servicing means lie completely within the coextensive boundaries of the plug, and when in the expanded position, some of the rigid support means and servicing means lie without the coextensive boundaries of the plug

  14. Advanced core monitoring technology for WWER reactors

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Casadei, A.L.; Doshi, P.K.

    1993-01-01

    The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

  15. Research on plasma core reactors

    International Nuclear Information System (INIS)

    Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

    1977-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF 6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm 3 aluminum canister in the central region was fueled with UF 6 gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation

  16. Restraint system for core elements of a reactor core

    International Nuclear Information System (INIS)

    Class, G.

    1975-01-01

    In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

  17. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Lai Shigang; Sun Libin; Zhang Zhengming

    2013-01-01

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  18. Method for refuelling a nuclear reactor core

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This invention relates to an improved method for refuelling a nuclear reactor core inside a reactor vessel. The technique allows a substantial reduction in the refuelling time as compared with previously known methods and permits fewer out of core operations and smaller temporary storage space. (U.K.)

  19. Reactor core design aiding system

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Hamaguchi, Yukio; Nakao, Takashi; Kondo, Yasuhide

    1995-01-01

    A two-dimensional radial power distribution and an axial one-dimensional power distribution are determined based on a distribution of a three-dimensional infinite multiplication factor, to obtain estimated power distribution estimation values. The estimation values are synthesized to obtain estimated three-dimensional power distribution values. In addition, the distribution of a two-dimensional radial multiplication factor and the distribution of an one-dimensional axial multiplication factor are determined based on the three-dimensional power distribution, to obtain estimated values for the multiplication factor distribution. The estimated values are synthesized to form estimated values for the three-dimensional multiplication factor distribution. Further, estimated fuel loading pattern value is determined based on the three-dimensional power distribution or the two-dimensional radial power distribution. Since the processes for determining the estimated values comprise only additive and multiplying operations, processing time can be remarkably saved compared with calculation based on a detailed physical models. Since the estimation is performed on every fuel assemblies, a nervous circuit network not depending on the reactor core system can be constituted. (N.H.)

  20. Solid oxide electrochemical reactor science.

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, Neal P. (Colorado School of Mines, Golden, CO); Stechel, Ellen Beth; Moyer, Connor J. (Colorado School of Mines, Golden, CO); Ambrosini, Andrea; Key, Robert J. (Colorado School of Mines, Golden, CO)

    2010-09-01

    Solid-oxide electrochemical cells are an exciting new technology. Development of solid-oxide cells (SOCs) has advanced considerable in recent years and continues to progress rapidly. This thesis studies several aspects of SOCs and contributes useful information to their continued development. This LDRD involved a collaboration between Sandia and the Colorado School of Mines (CSM) ins solid-oxide electrochemical reactors targeted at solid oxide electrolyzer cells (SOEC), which are the reverse of solid-oxide fuel cells (SOFC). SOECs complement Sandia's efforts in thermochemical production of alternative fuels. An SOEC technology would co-electrolyze carbon dioxide (CO{sub 2}) with steam at temperatures around 800 C to form synthesis gas (H{sub 2} and CO), which forms the building blocks for a petrochemical substitutes that can be used to power vehicles or in distributed energy platforms. The effort described here concentrates on research concerning catalytic chemistry, charge-transfer chemistry, and optimal cell-architecture. technical scope included computational modeling, materials development, and experimental evaluation. The project engaged the Colorado Fuel Cell Center at CSM through the support of a graduate student (Connor Moyer) at CSM and his advisors (Profs. Robert Kee and Neal Sullivan) in collaboration with Sandia.

  1. Lateral restraint assembly for reactor core

    Science.gov (United States)

    Gorholt, Wilhelm; Luci, Raymond K.

    1986-01-01

    A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

  2. Core monitoring at the WNP-2 reactor

    International Nuclear Information System (INIS)

    Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

    1992-01-01

    The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

  3. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

    Brown, W.L.; Geronime, R.L.

    1978-01-01

    Sensors including radiation detectors and the like for use within the core of nuclear reactors and which are constructed in a manner to provide optimum reliability of the sensor during use are described

  4. Core catcher for nuclear reactor core meltdown containment

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Bowman, F.L.

    1978-01-01

    A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

  5. Developments in gaseous core reactor technology

    International Nuclear Information System (INIS)

    Diaz, N.J.; Dugan, E.T.

    1979-01-01

    An effort to characterize the most promising concepts for large, central-station electrical generation was done under the auspices of the Nonproliferation Alternative Systems Assessment Program (NASAP). The two leading candidates were identified from this effort: The Mixed-Flow Gaseous Core Reactor (MFGCR) and the Heterogeneous Gas Core Reactor (HGCR). Key advantages over other nuclear concepts are weighed against the disadvantages of an unproven technology and the cost-time for deployment to make a sound decision on RandD support for these promising reactor alternatives. 38 refs

  6. Core homogenization method for pebble bed reactors

    International Nuclear Information System (INIS)

    Kulik, V.; Sanchez, R.

    2005-01-01

    This work presents a core homogenization scheme for treating a stochastic pebble bed loading in pebble bed reactors. The reactor core is decomposed into macro-domains that contain several pebble types characterized by different degrees of burnup. A stochastic description is introduced to account for pebble-to-pebble and pebble-to-helium interactions within a macro-domain as well as for interactions between macro-domains. Performance of the proposed method is tested for the PROTEUS and ASTRA critical reactor facilities. Numerical simulations accomplished with the APOLLO2 transport lattice code show good agreement with the experimental data for the PROTEUS reactor facility and with the TRIPOLI4 Monte Carlo simulations for the ASTRA reactor configuration. The difference between the proposed method and the traditional volume-averaged homogenization technique is negligible while only one type of fuel pebbles present in the system, but it grows rapidly with the level of pebble heterogeneity. (authors)

  7. Device for protecting deformations of reactor cores

    International Nuclear Information System (INIS)

    Kato, Yasuyoshi; Urushihara, Hiroshi.

    1975-01-01

    Object: To provide a fluid pressure cylinder, which is operated according to change in temperature of coolant for a reactor to restrain or release a core, to simply and effectively protect deformation of the core. Structure: A closed fluid pressure cylinder interiorly filled with suitable fluid is disposed in peripherally equally spaced relation in an annular space between a core barrel of a reactor and a reactor vessel. A piston is mounted in fluid-tight fashion in a plurality of piston openings made in the cylinder, the piston being slidably moved according to expansion and contraction of the fluid filled in the cylinder. The piston has a movable frame mounted at the foremost end thereof, the movable frame being moved integral with the piston, and the surface opposite the mount thereof biasing the outermost peripheral surface of the core. (Kamimura, M.)

  8. Intrinsically secure fast reactors with dense cores

    International Nuclear Information System (INIS)

    Slessarev, Igor

    2007-01-01

    Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important 'painful points' of nuclear power. Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety 'strategy' to reveal this worth. Such strategy is elaborated focusing on the priority for intrinsically secure safety features as well as on sure protection being provided by the first barrier of defence. Concerning the potential of fast reactors (i.e. sodium cooled, lead-cooled, etc.), there are no doubts that they are able to possess many favourable intrinsically secure safety features and to lay the proper foundation for a new reactor generation. However, some of their neutronic characteristics have to be radically improved. Among intrinsically secure safety properties, the following core parameters are significantly important: reactivity margin values, reactivity feed-back and coolant void effects. Ways of designing intrinsically secure safety features in fast reactors (titled hereafter as Intrinsically Secure Fast Reactors - ISFR) can be found in the frame of current reactor technologies by radical enhancement of core neutron economy and by optimization of core compositions. Simultaneously, respecting resistance to proliferation, by using non-enriched fuel feed as well as a core breeding gain close to zero, are considered as the important features (long-lived waste problems will be considered in a separate paper). This implies using the following reactor design options as well as closed fuel cycles with natural U as the reactor feed: ·Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. ·Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total power. The modules can be used also independently (as

  9. Core baffle for nuclear reactors

    International Nuclear Information System (INIS)

    Machado, O.J.; Berringer, R.T.

    1977-01-01

    The invention concerns the design of the core of a LWR with a large number of fuel assemblies formed by fuel rods and kept in position by spacer grids. According to the invention, at the level of the spacer grids match plates are mounted with openings so the flow of coolant directed upwards will not be obstructed and a parallel bypass will be obtained in the space between the core barrel and the baffle plates. In case of an accident, this configuration reduces or avoids damage from overpressure reactions. (HP) [de

  10. Innovative reactor core: potentialities and design

    International Nuclear Information System (INIS)

    Artioli, C.; Petrovich, Carlo; Grasso, Giacomo

    2010-01-01

    Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GW e y. (authors)

  11. Validation of reactor core protection system

    International Nuclear Information System (INIS)

    Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

    2008-01-01

    Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

  12. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  13. Reactor Structure Materials: Corrosion of Reactor Core Internals

    International Nuclear Information System (INIS)

    Van Dyck, S.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. Core construction for nuclear reactors

    International Nuclear Information System (INIS)

    Pettinger, D.S.

    1977-01-01

    HTR core construction with prismatic graphite blocks piled into columns. The front of the blocks is concavely curved. The lines of contact of two blocks are always not vertical, i.e. the blocks of one column are supported by the blocks of neighbouring columns so that ducts are formed. Groups of three or four of these columns may additionally be arranged around a central column which has recesses in order to lock the blocks of one group together. With this arrangement, dimensional changes of the graphite blocks under operating conditions can be taken up. (DG) [de

  16. Fluidized Bed Reactor as Solid State Fermenter

    Directory of Open Access Journals (Sweden)

    Krishnaiah, K.

    2005-01-01

    Full Text Available Various reactors such as tray, packed bed, rotating drum can be used for solid-state fermentation. In this paper the possibility of fluidized bed reactor as solid-state fermenter is considered. The design parameters, which affect the performances are identified and discussed. This information, in general can be used in the design and the development of an efficient fluidized bed solid-state fermenter. However, the objective here is to develop fluidized bed solid-state fermenter for palm kernel cake conversion into enriched animal and poultry feed.

  17. Heterogeneous cores for fast breeder reactor

    International Nuclear Information System (INIS)

    Schroeder, R.; Spenke, H.

    1980-01-01

    Firstly, the motivation for heterogeneous cores is discussed. This is followed by an outline of two reactor designs, both of which are variants of the combined ring and island core. These designs are presented by means of figures and detailed tables. Subsequently, a description of two international projects at fast critical zero energy facilities is given. Both of them support the nuclear design of heterogeneous cores. In addition to a survey of these projects, a typical experiment is discussed: the measurement of rate distributions. (orig.) [de

  18. Reactor core flow rate control system

    International Nuclear Information System (INIS)

    Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

    1996-01-01

    When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

  19. Support structure for reactor core constituent element

    International Nuclear Information System (INIS)

    Aida, Yasuhiko.

    1993-01-01

    A connection pipe having an entrance nozzle inserted therein as a reactor core constituent element is protruded above the upper surface of a reactor core support plate. A through hole is disposed to the protruding portion of the connection pipe. The through hole and a through hole disposed to the reactor core support plate are connected by a communication pipe. A shear rod is disposed in a horizontal portion at the inside of the communication pipe and is supported by a spring horizontally movably. Further, a groove is disposed at a position of the entrance nozzle opposing to the shear rod. The shear rod is urged out of the communication pipe by the pressure of the high pressure plenum and the top end portion of the shear rod is inserted to the groove of the entrance nozzle during operation. Accordingly, the shear rod is positioned in a state where it is extended from the through hole of the communication pipe to the groove of the entrance nozzle. This can mechanically constrain the rising of the reactor core constituent elements by the shear rod upon occurrence of earthquakes. (I.N.)

  20. Gas core reactors for coal gasification

    International Nuclear Information System (INIS)

    Weinstein, H.

    1976-01-01

    The concept of using a gas core reactor to produce hydrogen directly from coal and water is presented. It is shown that the chemical equilibrium of the process is strongly in favor of the production of H 2 and CO in the reactor cavity, indicating a 98 percent conversion of water and coal at only 1500 0 K. At lower temperatures in the moderator-reflector cooling channels the equilibrium strongly favors the conversion of CO and additional H 2 O to CO 2 and H 2 . Furthermore, it is shown the H 2 obtained per pound of carbon has 23 percent greater heating value than the carbon so that some nuclear energy is also fixed. Finally, a gas core reactor plant floating in the ocean is conceptualized which produces H 2 , fresh water and sea salts from coal

  1. Device for supporting a nuclear reactor core

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The core of a light-water reactor which is enclosed in a prestressed concrete pressure vessel and held within a diffuser basket is supported by a device consisting of a cylindrical shell which surrounds the basket and is rigidly fixed to a plurality of frusto-conical skirts having concurrent axes and located substantially at right angles to the axis of the reactor core. The small base of each skirt is rigidly fixed to the shell and the large base is anchored in openings formed in the reactor vessel for the penetration of coolant inlet and outlet pipes. The top portion of the shell is secured to the top portion of the diffuser basket, a flat surface being formed on the shell at the point of connection with each frusto-conical skirt so as to ensure rigid suspension while permitting thermal expansion

  2. 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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  4. Nuclear reactor core modelling in multifunctional simulators

    International Nuclear Information System (INIS)

    Puska, E.K.

    1999-01-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  5. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  6. Core clamping device for a nuclear reactor

    International Nuclear Information System (INIS)

    Guenther, R.W.

    1974-01-01

    The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

  7. Neutronic Core Performance of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel; Matzkin, S

    2000-01-01

    The actual design state of core of CAREM-25 reactor is presented.It is shown that the core design complains with the safety and operation established requirements.It is analyzed the behavior of the reactor safety and control systems (single failure of the fast shut down system, single failure of the shut down system, single failure of the second shut down system, reactivity worth of the adjust and control system in normal operation and hot shut down, reactivity worth of the adjust and control system and the scheme of movement of the control rod during the operation cycle).It is shown the burnup profile of fuel elements with the proposed scheme of refueling and the burnup and power density distribution at different moments of the operation cycle.The power peaking factor of the equilibrium core is 2.56, the minimum DNBR is 1.90 and its average is 2.09 during the operation cycle

  8. Apparatus for simulating a reactor core

    International Nuclear Information System (INIS)

    Yokomizo, Osamu; Kiguchi, Takashi; Motoda, Hiroshi; Takeda, Renzo.

    1975-01-01

    Object: To facilitate searching of input and output of information and to efficiently perform trial-and-error in a short time. Structure: Kinds of necessary input information are stored in an input information converter and are displayed by an image display through an image control. An operator operates an information input device to input information. This input information is converted by the input information converter into a form used in a reactor core simulation counter. The reactor core simulation counter simulates a state of the core to the input information converted, and outputs it as an output information. An output information converter converts output information into a form that may be displayed as an image and feeds it to the image control. The operator may correct the input information while viewing the output information displayed on the image display to immediately perform succeeding calculation. (Kamimura, M.)

  9. Conceptual core model for the reactor core test

    International Nuclear Information System (INIS)

    Swenson, L.D.

    1970-01-01

    Several design options for the ZrH Flight System Reactor were investigated which involved tradeoffs of core excess reactivity, reactor control, coolant mixing and cladding thickness. A design point was selected which is to be the basis for more detailed evaluation in the preliminary design phase. The selected design utilizes 295 elements with 0.670 inch element-to-element pitch, 32 mil thick Incoloy cladding, 18.00 inches long fuel meat, hydrogen content of 6.3 x 10 22 atoms/cc fuel, 10.5 w/o uranium, and a spiraled fin configuration with alternate elements having fins with spiral to the right, spiral to the left, and no fin at all (R-L-N fin configuration). Fin height is 30 mils for the center region of the core and 15 mils for the outer region. (U.S.)

  10. Device for removing a spent reactor core instrument tube

    International Nuclear Information System (INIS)

    Watanabe, Shigeru; Tsuji, Teruaki.

    1980-01-01

    Purpose: To easily and exactly execute works for removing a used reactor core instrument tube to be mounted in a reactor core from the lattice space of the core or for charging the tube into the lattice of the core. Constitution: When fuel assembly is pulled out of a reactor core and a spent reactor core instrument tube is then bent and removed from the core at periodical inspection time, a lower gripping unit integral with an upper gripping unit and a bending unit is provided at the lower end of a hanging rope of a winch, and lowered to the reactor core. Then, the spent reactor core instrument tube is gripped by the upper and lower gripping units, the bending unit is operated, the spent reactor core instrument tube is bent, and the tube is then pulled upwardly by the winch to remove the tube. (Aizawa, K.)

  11. Mechanical core coupling and reactors stability

    International Nuclear Information System (INIS)

    Suarez Antola, R.

    2006-01-01

    Structural parts of nuclear reactors are complex mechanical systems, able to vibrate with a set of proper frequencies when suitably excited. Cyclical variations in the strain state of the materials, including density perturbations, are produced. This periodic changes may affect reactor reactivity. But a variation in reactivity affects reactor thermal power, thus modifying the temperature field of the abovementiones materials. If the variation in temperature fields is fast enough, thermal-mechanical coupling may produce fast variations in strain states, and this, at its turn, modifies the reactivity, and so on. This coupling between mechanical vibrations of the structure and the materials of the core, with power oscillations of the reactor, not only may not be excluded a priori, but it seems that it has been present in some stage of the incidents or accidents that happened during the development of nuclear reactor technology. The purpose of the present communication is: (a) To review and generalize some mathematical models that were proposed in order to describe thermal-mechanical coupling in nuclear reactors. (b) To discuss some conditions in which significant instabilities could arise, including large amplitude power oscillations coupled with mechanical vibrations whose amplitudes are too small to be excluded by conventional criteria of mechanical design. Enough Certain aspects of thr physical safety of nuclear power reactors, that are objected by people that opposes to the renaissance of nucleoelectric generation, are discussed in the framework of the mathematical model proposed in this paper [es

  12. A Large Solid Inner Core at Mercury

    Science.gov (United States)

    Genova, A.; Goossens, S.; Mazarico, E.; Lemoine, F. G.; Neumann, G. A.; Kuang, W.; Sabaka, T. J.; Smith, D. E.; Zuber, M. T.

    2018-05-01

    New measurements of the polar moments of inertia of the whole planet and of the outer layers (crust+mantle), and simulations of Mercury’s magnetic field dynamo suggest the presence of a solid inner core with a Ric 0.3-0.5 Roc.

  13. Overheating preventive system for reactor core fuels

    International Nuclear Information System (INIS)

    Ito, Daiju

    1981-01-01

    Purpose: To ensure the cooling function of reactor water in a cooling system in case of erroneous indication or misoperation by reliable temperature measurement for fuels and actuating relays through the conversion output obtained therefrom. Constitution: Thermometers are disposed laterally and vertically in a reactor core in contact with core fuels so as to correspond to the change of status in the reactor core. When there is a high temperature signal issued from one of the thermometers or one of conversion circuits, the function of relay contacts does not provide the closed state as a whole. When high temperature signals are issued from two or more thermometers of conversion circuits from independent OR circuits, the function of the relay contacts provides the closure state as a whole. Consequently, in the use of 2-out of 3-circuits, the entire closure state, that is, the misoperation of the relay contacts for the thermometer or the conversion circuits can be avoided. In this way, by the application of the output from the conversion circuits to the logic circuit and, in turn, application of the output therefrom to the relay groups in 2-out of 3-constitution, the reactor safety can be improved. (Horiuchi, T.)

  14. Fuel assembly and nuclear reactor core

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

    1995-01-01

    The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

  15. Neutronics of a mixed-flow gas-core reactor

    International Nuclear Information System (INIS)

    Soran, P.D.; Hansen, G.E.

    1977-11-01

    The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

  16. Liquid metal reactor core material HT9

    International Nuclear Information System (INIS)

    Kim, S. H.; Kuk, I. H.; Ryu, W. S. and others

    1998-03-01

    A state-of-the art is surveyed on the liquid metal reactor core materials HT9. The purpose of this report is to give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is planned for the year of 2010. In-core stability of cladding materials is important to the extension of fuel burnup. Austenitic stainless steel (AISI 316) has been used as core material in the early LMR due to the good mechanical properties at high temperatures, but it has been found to show a poor swelling resistance. So many efforts have been made to solve this problem that HT9 have been developed. HT9 is 12Cr-1MoVW steel. The microstructure of HT9 consisted of tempered martensite with dispersed carbide. HT9 has superior irradiation swelling resistance as other BCC metals, and good sodium compatibility. HT9 has also a good irradiation creep properties below 500 dg C, but irradiation creep properties are degraded above 500 dg C. Researches are currently in progress to modify the HT9 in order to improve the irradiation creep properties above 500 dg C. New design studies for decreasing the core temperature below 500 dg C are needed to use HT9 as a core material. On the contrary, decrease of the thermal efficiency may occur due to lower-down of the operation temperature. (author). 51 refs., 6 tabs., 19 figs

  17. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  18. Graphite core design in UK reactors

    International Nuclear Information System (INIS)

    Davies, M.W.

    1996-01-01

    The cores in the first power producing Magnox reactors in the UK were designed with only a limited amount of information available regarding the anisotropic dimensional change behaviour of Pile Grade graphite. As more information was gained it was necessary to make modifications to the design, some minor, some major. As the cores being built became larger, and with the switch to the Advanced Gas-cooled Reactor (AGR) with its much higher power density, additional problems had to be overcome such as increased dimensional change and radiolytic oxidation by the carbon dioxide coolant. For the AGRs a more isotropic graphite was required, with a lower initial open pore volume and higher strength. Gilsocarbon graphite was developed and was selected for all the AGRs built in the UK. Methane bearing coolants are used to limit radiolytic oxidation. (author). 5 figs

  19. PC-Reactor-core transient simulation code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author) [pt

  20. Core access system for nuclear reactor

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved nuclear reactor arrangement to facilitate both through-the-head instrumentation and insertion and removal of assemblies from the nuclear core. The arrangement is of the type including a reactor vessel head comprising a large rotatable cover having a plurality of circular openings therethrough, a plurality of upwardly extending nozzles mounted on the upper surface of a large cover, and a plurality of upwardly extending skirts mounted on a large cover about the periphery or boundary of the circular openings; a plurality of small plugs for each of the openings in the large cover, the plugs also having nozzles mounted on the upper surface thereof, and drive mechanisms mounted on top of some of the nozzles and having means extending therethrough into the reactor vessel, the drive mechanisms and nozzles extending above the elevation of the upwardly extending skirts

  1. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  2. Nuclear reactor core and fuel element therefor

    International Nuclear Information System (INIS)

    Fortescue, P.

    1986-01-01

    This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces

  3. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  4. Solving the uncommon reactor core neutronics problems

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1980-01-01

    The common reactor core neutronics problems have fundamental neutron space, energy spectrum solutions. Typically the most positive eigenvalue is associated with an all-positive flux for the pseudo-steady-state condition (k/sub eff/), or the critical state is to be effected by selective adjustment of some variable such as the fuel concentration. With sophistication in reactor analysis has come the demand for solutions of other, uncommon neutronics problems. Importance functionss are needed for sensitivity and uncertainty analyses, as for ratios of intergral reaction rates such as the fuel conversion (breeding) ratio. The dominant higher harmonic solution is needed in stability analysis. Typically the desired neutronics solution must contain negative values to qualify as a higher harmonic or to satisfy a fixed source containing negative values. Both regular and adjoint solutions are of interest as are special integrals of the solutions to support analysis

  5. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

    1980-01-01

    A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter. (UK)

  6. Sensors for use in nuclear reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-21

    A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter.

  7. Core of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

    Giacometti, Christian; Mougniot, J.-C.; Ravier, Jean.

    1974-01-01

    The fast neutron nuclear reactor described includes an internal area in fissile material completely enclosed in an area of fertile material forming the outside blanket. The internal fissile area is provided with housings exclusively filled with fertile material forming one or more inside blankets. In this core the internal blankets are shaped like rings vertically separating superimposed rings of fissile material. The blanket of material nearest to the periphery is circumscribed externally by a contour having an indented shape on its straight section so as to increase the contact area between this blanket and the external blanket [fr

  8. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  9. In core system mapping reactor power distribution

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Moreira, J.M.L.

    1989-01-01

    Based on the signals of SPND'S (Self Powered Neutron Detectors) distributed inside of a core, the spatial power distribution is obtained using the MAP program, developed in this work. The methodology applied in MAP program uses a least mean square technique to calculate expansion coefficients that depend on the SPND'S signals. The final power or neutron flux distribution is obtained by a combination of certains functions or expansion modes that are provided from diffusion calculation with the CITATION code. The MAP program is written in PASCAL language and will be used in IEA-R1 reactor for assisting its operation. (author) [pt

  10. Statistical hot spot analysis of reactor cores

    International Nuclear Information System (INIS)

    Schaefer, H.

    1974-05-01

    This report is an introduction into statistical hot spot analysis. After the definition of the term 'hot spot' a statistical analysis is outlined. The mathematical method is presented, especially the formula concerning the probability of no hot spots in a reactor core is evaluated. A discussion with the boundary conditions of a statistical hot spot analysis is given (technological limits, nominal situation, uncertainties). The application of the hot spot analysis to the linear power of pellets and the temperature rise in cooling channels is demonstrated with respect to the test zone of KNK II. Basic values, such as probability of no hot spots, hot spot potential, expected hot spot diagram and cumulative distribution function of hot spots, are discussed. It is shown, that the risk of hot channels can be dispersed equally over all subassemblies by an adequate choice of the nominal temperature distribution in the core

  11. 78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2013-09-12

    ... 52 [Docket No. PRM-50-105; NRC-2012-0056] In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission. ACTION: Petition for rulemaking; denial...-core thermocouples at different elevations and radial positions throughout the reactor core to enable...

  12. Core arrangement in BWR type reactors

    International Nuclear Information System (INIS)

    Asano, Masayuki.

    1981-01-01

    Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

  13. Life extension of CANDU reactor cores

    International Nuclear Information System (INIS)

    Millard, J.; Kerker, J.; Albert, M.

    2011-01-01

    Candu Energy (formerly AECL), in partnership with station operators, has developed a robust methodology for demonstrating the fitness of reactor core structures, and associated reactivity control devices, as an essential element in conducting a station life extension project. The ageing of reactors is affected by ageing mechanisms impacted by operational history and design related factors such as materials, chemistries and stress distributions. The methodology of this life extension work is based on the IAEA TECDOC 1197; which documents practices for ageing management in CANDU reactors. This paper uses the work in Bruce Units 1 and 2, conducted from 2007 through to 2011, to explain the methodology. The work started with analysis of historical operational conditions and identification of the forms of degradation that could have occurred. The assessment and related inspections considered the safety and pressure boundary significance of each item, as well as its failure modes and margins. It then moved through both general and local inspection, focused mainly inside the calandria vessel once the calandria tubes were removed. The inspection found the bulk of the hardware to be in good condition, with a small number of remediation opportunities. In the course of that remediation some foreign material was sampled and removed. The minor remediation was successful and the work was completed through formal documentation of the fitness for extended life. It has been demonstrated through these analyses and visual inspections that the reactor structures and components inspected are free of indications and active degradation mechanisms that would prevent the safe and reliable operation of Bruce A Units 1 and 2 through its next 25 years of life. (author)

  14. Seismic response of a block-type nuclear reactor core

    International Nuclear Information System (INIS)

    Dove, R.C.; Bennett, J.G.; Merson, J.L.

    1976-05-01

    An analytical model is developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies

  15. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  16. Spring unit especially intended for a nuclear reactor core

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, Wilhelm.

    1977-01-01

    This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

  17. Reactor core of light water-cooled reactor

    International Nuclear Information System (INIS)

    Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

    1996-01-01

    In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

  18. Criteria design of the CAREM 25 reactor's core: neutronic aspects

    International Nuclear Information System (INIS)

    Lecot, C.A.

    1990-01-01

    The criteria that guided the design, from the neutronic point of view, of the CAREM reactor's core were presented. The minimum set of objectives and general criteria which permitted the design of the particular systems constituting the CAREM 25 reactor's core is detailed and stated. (Author) [es

  19. Neutronic analysis of the ford nuclear reactor leu core

    International Nuclear Information System (INIS)

    Raza, S.S.; Hayat, T.

    1989-08-01

    Neutronic analysis of the ford nuclear reactor low enriched uranium core has been carried out to gain confidence in the com puting methodology being used for Pakistan Research Reactor-1 core conversion calculations. The computed value of the effective multiplication factor (Keff) is found to be in good agreement with that quoted by others. (author). 6 figs

  20. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  1. Applications of plasma core reactors to terrestrial energy systems

    International Nuclear Information System (INIS)

    Lantham, T.S.; Biancardi, F.R.; Rodgers, R.J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrail applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times

  2. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

  3. The 2nd reactor core of the NS Otto Hahn

    International Nuclear Information System (INIS)

    Manthey, H.J.; Kracht, H.

    1979-01-01

    Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

  4. Gas-core reactor power transient analysis. Final report

    International Nuclear Information System (INIS)

    Kascak, A.F.

    1972-01-01

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

  5. Analysis of core and core barrel heat-up under conditions simulating severe reactor accidents

    International Nuclear Information System (INIS)

    Chellaiah, S.; Viskanta, R.; Ranganathan, P.; Anand, N.K.

    1987-01-01

    This paper reports on the development of a model for estimating the temperature distributions in the reactor core, core barrel, thermal shield and reactor pressure vessel of a PWR during an undercooling transient. A number of numerical calculations simulating the core uncovering of the TMI-2 reactor and the subsequent heat-up of the core have been performed. The results of the calculations show that the exothermic heat release due to Zircaloy oxidation contributes to the sharp heat-up of the core. However, the core barrel temperature rise which is driven by the temperature increase of the edge of the core (e.g., the core baffle) is very modest. The maximum temperature of the core barrel never exceeded 610 K (at a system pressure of 68 bar) after a 75 minute simulation following the start of core uncovering

  6. The seismic assessment of fast reactor cores in the UK

    International Nuclear Information System (INIS)

    Duthie, J.C.; Dostal, M.

    1988-01-01

    The design of the UK Commercial Demonstration Fast Reactor (CDFR) has evolved over a number of years. The design has to meet two seismic requirements: (i) the reactor must cause no hazard to the public during or after the Safe Shutdown Earthquake (SSE); (ii) there must be no sudden reduction in safety for an earthquake exceeding the SSE. The core is a complicated component in the whole reactor. It is usually represented in a very simplified manner in the seismic assessment of the whole reactor station. From this calculation, a time history or response spectrum can be generated for the diagrid, which supports the core, and for the above core structure, which supports the main absorber rods. These data may then be used to perform a detailed assessment of the reactor core. A new simplified model of the core response may then be made and used in a further calculation of the whole reactor. The calculation of the core response only, is considered in the remainder of this paper. One important feature of the fast reactor core, compared with other reactors, is that the components are relatively thin and flexible to promote neutron economy and heat transfer. A further important feature is that there are very small gaps between the wrapper tubes. This leads to very strong fluid-coupling effects. These effects are likely to be beneficial, but adequate techniques to calculate them are only just being developed. 9 refs, figs

  7. Measuring device for the coolant flowrate in a reactor core

    International Nuclear Information System (INIS)

    Sawa, Toshihiko.

    1983-01-01

    Purpose: To improve the operation performance by enabling direct and accurate measurement for the reactor core recycling flowrate. Constitution: A control rod guide is disposed to the upper end of a control rod drive mechanism housing passing through the bottom of a reactor pressure vessel and it is inserted into the through hole of a reactor core support plate. A water flow passage is formed through the reactor core support plate for the flowrate measurement of coolants recycled within the reactor core. The static pressure difference between the upper and the lower sides of the reactor core support plate is measured by a pressure difference detector of a pressure difference measuring mechanism, and an output signal from the pressure different detector is inputted to a calculation means, in which the amount of the coolants passing through the water flow passage is calculated based on the output signal corresponding to the pressure difference. Then, the total recycling flowrate in the reactor core is determined in the calculation means based on the relation between the measured flowrate and a predetermined total reactor core recycling flowrate. (Horiuchi, T.)

  8. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

    2009-01-01

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

  9. Development of high performance core for large fast breeder reactors

    International Nuclear Information System (INIS)

    Inoue, Kotaro; Kawashima, Katsuyuki; Watari, Yoshio.

    1982-01-01

    Subsequently to the fast breeder prototype reactor ''Monju'', the construction of a demonstration reactor with 1000 MWe output is planned. This research aims at the establishment of the concept of a large core with excellent fuel breeding property and safety for a demonstration and commercial reactors. For the purpose, the optimum specification of fuel design as a large core was clarified, and the new construction of a core was examined, in which a disk-shaped blanket with thin peripheral edge is introduced at the center of a core. As the result, such prospect was obtained that the time for fuel doubling would be 1/2, and the energy generated in a core collapse accident would be about 1/5 as compared with a large core using the same fuel as ''Monju''. Generally, as a core is enlarged, the rate of breeding lowers. If a worst core collapse accident occurs, the scale of accident will be very large in the case of a ''Monju'' type large core. In an unhomogeneous core, an internal blanket is provided in the core for the purpose of improving the breeding property and safety. Hitachi Ltd. developed the concept of a large core unhomogeneous in axial direction and proposed it. The research on the fuel design for a large core, an unhomogeneous core and its core collapse accident is reported. (Kako, I.)

  10. Lifetime embrittlement of reactor core materials

    International Nuclear Information System (INIS)

    Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G.; White, C.J.

    1994-08-01

    Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10 24 n/M 2 (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K IC due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K IC plus its lower hydrogen absorption characteristics compared with Zircaloy-2

  11. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

    2002-03-01

    In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

  12. Core design concepts for high performance light water reactors

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.

    2007-01-01

    Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

  13. Development of inherent core technologies for advanced reactor

    International Nuclear Information System (INIS)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H.

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  14. Development of inherent core technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  15. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  16. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  17. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  18. Advanced gadolinia core and Toshiba advanced reactor management system

    International Nuclear Information System (INIS)

    Miyamoto, Toshiki; Yoshioka, Ritsuo; Ebisuya, Mitsuo

    1988-01-01

    At the Hamaoka Nuclear Power Station, Unit No. 3, advanced core design and core management technology have been adopted, significantly improving plant availability, operability and reliability. The outstanding technologies are the advanced gadolinia core (AGC) which utilizes gadolinium for the axial power distribution control, and Toshiba advanced reactor management system (TARMS) which uses a three-dimensional core physics simulator to calculate the power distribution. Presented here are the effects of these advanced technologies as observed during field testing. (author)

  19. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  20. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

    The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

  1. Modeling of reflood of severely damaged reactor core

    International Nuclear Information System (INIS)

    Bachrata, A.

    2012-01-01

    The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

  2. Improving Battery Reactor Core Design Using Optimization Method

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2011-01-01

    The Battery Omnibus Reactor Integral System (BORIS) is a small modular fast reactor being designed at Seoul National University to satisfy various energy demands, to maintain inherent safety by liquid-metal coolant lead for natural circulation heat transport, and to improve power conversion efficiency with the Modular Optimal Balance Integral System (MOBIS) using the supercritical carbon dioxide as working fluid. This study is focused on developing the Neutronics Optimized Reactor Analysis (NORA) method that can quickly generate conceptual design of a battery reactor core by means of first principle calculations, which is part of the optimization process for reactor assembly design of BORIS

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

    International Nuclear Information System (INIS)

    Takiyama, Masashi.

    1994-01-01

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

  4. Fuel loading method to exchangeable reactor core of BWR type reactor and its core

    International Nuclear Information System (INIS)

    Koguchi, Kazushige.

    1995-01-01

    In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

  5. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  6. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  7. Methanol synthesis in a countercurrent gas-solid-solid trickle flow reactor. An experimental study

    NARCIS (Netherlands)

    Kuczynski, M.; Oyevaar, M.H.; Pieters, R.T.; Westerterp, K.R.

    1987-01-01

    The synthesis of methanol from CO and H2 was executed in a gas-solid-solid trickle flow reactor. The reactor consisted of three tubular reactor sections with cooling sections in between. The catalyst was Cu on alumina, the adsorbent was a silica-alumina powder and the experimental range 498–523 K,

  8. Stationary liquid fuel fast reactor SLFFR – Part I: Core design

    Energy Technology Data Exchange (ETDEWEB)

    Jing, T.; Yang, G.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

    2016-12-15

    Highlights: • An innovative fast reactor concept SLFFR based on liquid metal fuel is proposed for TRU burning. • A compact core design of 1000 MWt SLFFR is developed to achieve a zero conversion ratio and passive safety. • The core size and the control requirement are significantly reduced compared to the conventional solid fuel reactor with same conversion ratio. - Abstract: For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative.

  9. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    Anoussis, J.N.; Chrysochoides, N.G.; Papastergiou, C.N.

    1982-07-01

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  10. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  11. 77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2012-05-23

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [Docket No. PRM-50-105; NRC-2012-0056] In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission... of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples...

  12. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  13. Development of a core follow calculational system for research reactors

    International Nuclear Information System (INIS)

    Muller, E.Z.; Ball, G.; Joubert, W.R.; Schutte, H.C.; Stoker, C.C.; Reitsma, F.

    1994-01-01

    Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

  14. Neutron spectrometric methods for core inventory verification in research reactors

    International Nuclear Information System (INIS)

    Ellinger, A.; Filges, U.; Hansen, W.; Knorr, J.; Schneider, R.

    2002-01-01

    In consequence of the Non-Proliferation Treaty safeguards, inspections are periodically made in nuclear facilities by the IAEA and the EURATOM Safeguards Directorate. The inspection methods are permanently improved. Therefore, the Core Inventory Verification method is being developed as an indirect method for the verification of the core inventory and to check the declared operation of research reactors

  15. Core neutronics of a swimming pool research reactor

    International Nuclear Information System (INIS)

    Mannan, M.A.; Mondal, M.A.W.; Pervini, M.E.

    1981-01-01

    The initial cores of the 5 MW swimming pool research reactor of the Nuclear Research Centre, Tehran have been analyzed using the computer codes METHUSELAH and EQUIPOISE. The effective multiplication factor, critical mass, moderator temperature and void coefficients of the core have been calculated and compared with vendor's values. Calculated values agree reasonably well with the vendor's results. (author)

  16. Axial heterogeneous core concept applied for super phoenix reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

    1991-11-01

    Always maintaining the current design rules, this paper presents a parametric study on the type of axial heterogeneous core concept (CHA), utilizing a core of fast reactor Super Phenix type, reaching a maximum thermal burnup rate of 150000 M W d/t and being managed in single batch. (author)

  17. Fast reactor core concepts to improve transmutation efficiency

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-01-01

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate

  18. Research reactor core conversion guidebook. V.1: Summary

    International Nuclear Information System (INIS)

    1992-04-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs

  19. European ERANOS formulaire for fast reactor core analysis

    International Nuclear Information System (INIS)

    Rimpault, Gerald

    2003-01-01

    ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

  20. Safety characteristics of the US advanced liquid metal reactor core

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

    1991-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

  1. Supporting system for the core restraint of nuclear reactors

    International Nuclear Information System (INIS)

    Kaser, A.

    1973-01-01

    The core restraint of water cooled nuclear reactors which is needed to direct the flow of the coolant through the core can be manufactured only in a moderate wall thickness. Thus, the majority of the loads have to be transmitted to the core barrel which is more rigid. The patent refers to a system of circumferential and vertical support members most of which are free to move relatively to each other, thus reducing thermal stresses during operation. (P.K.)

  2. Characteristics of fast reactor core designs and closed fuel cycle

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  3. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    Gueray, B.S.

    2001-01-01

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  4. Thawing of lithium in the SP-100 reactor core configuration

    International Nuclear Information System (INIS)

    Magee, P.M.; Malovrh, J.W.; REineking, W.H.

    1986-01-01

    The General Electric SP-100 Liquid Metal Reactor is designed to be launched with the lithium coolant in the reactor and primary loops frozen. Initial startup of the system in space, after a satisfactory orbit is achieved, will be accomplished by slowly increasing the power in the reactor core and using the heat generated to melt the lithium, first in the reactor, and then progressively down the primary loops. This technique significantly facilitates ground handling, reduces vibrational loads during vehicle launch and minimized the shuttle bay heat load. The challenge is to thaw the coolant and startup the system within an acceptable time without structural damage. The test results clearly demonstrate that thawing of the lithium in the SP-100 reactor core can be done rapidly without structural damage and, thus, support the selected concept of SP-100 launch with frozen lithium and thaw/startup in space

  5. Simulating Neutronic Core Parameters in a Research and Test Reactor

    International Nuclear Information System (INIS)

    Selim, H.K.; Amin, E.A.; Koutb, M.E.

    2011-01-01

    The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload

  6. Radionuclide distribution in TMI-2 reactor building basement liquids and solids

    International Nuclear Information System (INIS)

    Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

    1984-01-01

    As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

  7. Vessel core seismic interaction for a fast reactor

    International Nuclear Information System (INIS)

    Martelli, A.; Maresca, G.

    1984-01-01

    This report deals with the analysis carried out in collaboration between ENEA and NIRA for optimizing the iterative procedure applied for the evaluation of the effects of the vessel core dynamic interaction for a fast reactor in the case of a earthquake. In fact, as shown in a previous report the convergence of such procedure was very slow for the design solution adopted for the PEC reactor, i.e. with a core restraint plate located close to the top of the core elements. This study, although performed making use of preliminary data (the same of the cited previous report) demonstrates that the convergence is fast if a suitable linear core model is applied in the first iteration linear calculations carried out by NIRA, with an intermediate stiffness with respect to those corresponding to the two limit models previously assumed and increased damping coefficients. Thus, the optimized iterative procedures is now applied in the PEC reactor block seismic verification analysis

  8. COMSORS: A light water reactor chemical core catcher

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.

    1997-01-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B 2 O 3 ) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

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

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

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

  10. Lunar Fluid Core and Solid-Body Tides

    Science.gov (United States)

    Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.

    2005-01-01

    Variations in rotation and orientation of the Moon are sensitive to solid-body tidal dissipation, dissipation due to relative motion at the fluid-core/solid-mantle boundary, and tidal Love number k2 [1,2]. There is weaker sensitivity to flattening of the core-mantle boundary (CMB) [2-5] and fluid core moment of inertia [1]. Accurate Lunar Laser Ranging (LLR) measurements of the distance from observatories on the Earth to four retroreflector arrays on the Moon are sensitive to lunar rotation and orientation variations and tidal displacements. Past solutions using the LLR data have given results for dissipation due to solid-body tides and fluid core [1] plus Love number [1-5]. Detection of CMB flattening has been improving [3,5] and now seems significant. This strengthens the case for a fluid lunar core.

  11. Actively doped solid core Photonic Bandgap Fiber

    DEFF Research Database (Denmark)

    Broeng, Jes; Olausson, Christina Bjarnal Thulin; Lyngsøe, Jens Kristian

    2010-01-01

    Solid photonic bandgap fibers offer distributed spectral filtering with extraordinary high suppression. This opens new possibilities of artificially tailoring the gain spectrum of fibers. We present record-performance of such fibers and outline their future applications....

  12. Fuel requirements for experimental devices in MTR reactors. A perturbation model for reactor core analysis

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1991-01-01

    Irradiation in neutron absorbing devices, requiring high fast neutron fluxes in the core or high thermal fluxes in the reflector and flux traps, lead to higher density fuel and larger core dimensions. A perturbation model of the reactor core helps to estimate the fuel requirements. (orig.)

  13. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  14. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  15. UV patterned nanoporous solid-liquid core waveguides

    DEFF Research Database (Denmark)

    Gopalakrishnan, Nimi; Sagar, Kaushal Shashikant; Christiansen, Mads Brøkner

    2010-01-01

    Nanoporous Solid-Liquid core waveguides were prepared by UV induced surface modification of hydrophobic nanoporous polymers. With this method, the index contrast (delta n = 0.20) is a result of selective water infiltration. The waveguide core is defined by UV light, rendering the exposed part...

  16. MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

    International Nuclear Information System (INIS)

    Relloso, J.M.

    1990-01-01

    This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author) [es

  17. Rotating solid foam reactors : mass transfer and reaction rate

    NARCIS (Netherlands)

    Tschentscher, R.

    2012-01-01

    In this thesis the performance and applicability of rotating solid foam stirrers is investigated. The stirrer consists, thereby of a solid, highly porous structure, which is used as stirrer and catalyst support simultaneously. The solid foam block occupies a large part of the reactor volume.

  18. Shock loading of reactor vessel following hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Srinivas, G.; Doshi, J.B.

    1990-01-01

    Hypothetical Core Disruptive Accident (HCDA) has been historically considered as the maximum credible accident in Fast Breeder Reactor systems. Environmental consequences of such an accident depends to a great extent on the ability of the reactor vessel to maintain integrity during the shock loading following an HCDA. In the present paper, a computational model of the reactor core and the surrounding coolant with a free surface is numerical technique. The equations for conservation of mass, momentum and energy along with an equation of state are considered in two dimensional cylindrical geometry. The reactor core at the end of HCDA is taken as a bubble of hot, vaporized fuel at high temperature and pressure, formed at the center of the reactor vessel and expanding against the surrounding liquid sodium coolant. The free surface of sodium at the top of the vessel and the movement of the core bubble-liquid coolant interface are tracked by Marker and Cell (MAC) procedure. The results are obtained for the transient pressure at the vessel wall and also for the loading on the roof plug by the impact of the slug of liquid sodium. The computer code developed is validated against a benchmark experiment chosen to be ISPRA experiment reported in literature. The computer code is next applied to predict the loading on the Indian Prototype Fast Breeder Reactor (PFBR) being developed at Kalpakkam

  19. Study on core design for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Okubo, Tsutomu

    2002-01-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  20. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  1. Molten core material holding device in a nuclear reactor

    International Nuclear Information System (INIS)

    Nakamura, Hisashi; Tanaka, Nobuo; Takahashi, Katsuro.

    1985-01-01

    Purpose: To improve the function of cooling to hold molten core materials in a molten core material holding device. Constitution: Plenum structures are formed into a pan-like configuration, in which liners made of metal having high melting point and relatively high heat conductivity such as tantalum, tungsten, rhenium or alloys thereof are integrally appended to hold and directly cool the molten reactor core materials. Further, a plurality of heat pipes, passing through the plenum structures, facing the cooling portion thereof to the coolants at the outer side and immersing the heating portion into the molten core materials fallen to deposit in the inner liners are disposed radially. Furthermore, heat pipes embodded in the plenum structure are disposed in the same manner below the liners. Thus, the plenum structures and the molten reactor core materials can be cooled at a high efficiency. (Seki, T.)

  2. 3D computer visualization and animation of CANDU reactor core

    International Nuclear Information System (INIS)

    Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

    1999-01-01

    Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

  3. Thermal barrier and support for nuclear reactor fuel core

    International Nuclear Information System (INIS)

    Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

    1987-01-01

    A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

  4. Automated Design and Optimization of Pebble-bed Reactor Cores

    International Nuclear Information System (INIS)

    Gougar, Hans D.; Ougouag, Abderrafi M.; Terry, William K.

    2010-01-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  5. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  6. Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

    International Nuclear Information System (INIS)

    Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

    2001-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

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

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

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

  8. An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Clark, J.S.; Walton, J.T.; Mcguire, M.L.

    1992-07-01

    Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines. 11 refs

  9. Core management and reactor physics aspects of the conversion of the NRU reactor to LEU

    International Nuclear Information System (INIS)

    Atfield, M.D.

    1985-01-01

    Results of work done to assess the effects of converting the NRU reactor to LEU are presented. The effects are small, and the operational rules and safety analysis, appropriate to the HEU core, will still apply. (author)

  10. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2016-09-15

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  11. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    International Nuclear Information System (INIS)

    Cuevas V, G.; Banfield, J.; Avila N, A.

    2016-09-01

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  12. Design and development of small and medium integral reactor core

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

    1997-07-01

    Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR's, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs

  13. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    Bonacina, G.; Castoldi, A.; Zola, M.; Cecchini, F.; Martelli, A.; Vincenzi, D.

    1982-01-01

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  14. A study of passive safety conditions for fast reactor core

    International Nuclear Information System (INIS)

    Shimizu, Akinao

    1991-01-01

    A study has been made for passive safety conditions of fast reactor cores. Objective of the study is to develop a concept of a core with passive safety as well as a simple safety philosophy. A simple safety philosophy, which is wore easy to explain to the public, is needed to enhance the public acceptance for nuclear reactors. The present paper describes a conceptual plan of the study including the definition of the problem a method of approach and identification of tasks to be solved

  15. Site Investigation for Detection of KIJANG Reactor Core Center

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Tae-Hyun; Kim, Jun Yeon; Kim, Jeeyoung [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    It was planned for the end of March 2017 and extended to April 2018 according to the government budget adjustment. The KJRR project is intended for filling the self-sufficiency of RI demand including Mo-99, increasing the NTD capacity and developing technologies related to the research reactor. In project, site investigation is the first activity that defines seismologic and related geologic aspects of the site. Site investigation was carried out from Oct. 2012 to Jan. 2014 and this study is intended to describe detail procedures in locating the reactor core center. The location of the reactor core center was determined by collectively reviewing not only geological information but also information from architects engineering. EL 50m was selected as ground level by levering construction cost. Four recommended locations (R-1a - R-1d) are displayed for the reactor core center. R-1a was found optimal in consideration of medium rock contour, portion of medium rock covering reactor buildings, construction cost, physical protection and electrical resistivity. It is noted that engineering properties of the medium rock is TCR/RQD 100/53, elastic modulus 7,710 - 8,720MPa, permeability coefficient 2.92E-06cm/s, and S-wave velocity 1,380m/s, sound for foundations of reactor buildings.

  16. Refurbishment, core conversion and safety analysis of Apsara reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K.; Sasidharan, K.; Sengupta, S. [Bhabha Atomic Research Centre, Mumbai (India)]. E-mail: nram@@apsara.barc.ernet.in

    1998-07-01

    Apsara, a 1 MWt pool type reactor using HEU fuel has been in operation at the Bhabha Atomic Research Centre, Trombay since 1956. In view of the long service period seen by the reactor it is now planned to carry out extensive refurbishment of the reactor with a view to extend its useful life. It is also proposed to modify the design of the reactor wherein the core will be surrounded by a heavy water reflector tank to obtain a good thermal neutron flux over a large radial distance from the core. Beam holes and the majority of the irradiation facilities will be located inside the reflector tank. The coolant flow direction through the core will be changed from the existing upward flow to downward flow. A delay tank, located inside the pool, is provided to facilitate decay of short lived radioactivity in the coolant outlet from the core in order to bring down radiation field in the operating areas. Analysis of various anticipated operational occurrences and accident conditions like loss of normal power, core coolant flow bypass, fuel channel blockage and degradation of primary coolant pressure boundary have been performed for the proposed design. Details of the proposed design modifications and the safety analyses are given in the paper. (author)

  17. The temperature distribution in a gas core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C. (Interuniversitair Reactor Inst., Delft (Netherlands)); Kistemaker, J.; Boersma-Klein, W.; Vitalis, F. (FOM-Instituut voor Atoom-en Molecuulfysica, Amsterdam (Netherlands))

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author).

  18. The temperature distribution in a gas core fission reactor

    International Nuclear Information System (INIS)

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

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author)

  19. Reconstruction calculation of pin power for ship reactor core

    International Nuclear Information System (INIS)

    Li Haofeng; Shang Xueli; Chen Wenzhen; Wang Qiao

    2010-01-01

    Aiming at the limitation of the software that pin power distribution for ship reactor core was unavailable, the calculation model and method of the axial and radial pin power distribution were proposed. Reconstruction calculations of pin power along axis and radius was carried out by bicubic and bilinear interpolation and cubic spline interpolation, respectively. The results were compared with those obtained by professional reactor physical soft with fine mesh difference. It is shown that our reconstruction calculation of pin power is simple and reliable as well as accurate, which provides an important theoretic base for the safety analysis and operating administration of the ship nuclear reactor. (authors)

  20. Core damage frequency (reactor design) perspectives based on IPE results

    International Nuclear Information System (INIS)

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

    1996-01-01

    This paper provides perspectives gained from reviewing 75 Individual Plant Examination (IPE) submittals covering 108 nuclear power plant units. Variability both within and among reactor types is examined to provide perspectives regarding plant-specific design and operational features, and C, modeling assumptions that play a significant role in the estimates of core damage frequencies in the IPEs. Human actions found to be important in boiling water reactors (BWRs) and in pressurized water reactors (PWRs) are presented and the events most frequently found important are discussed

  1. Constitutive relations for nuclear reactor core materials

    International Nuclear Information System (INIS)

    Zaverl, F. Jr.; Lee, D.

    1978-01-01

    A strain rate dependent constitutive equation is proposed which is capable of describing inelastic deformation behavior of anisotropic metals, such as Zircaloys, under complex loading conditions. The salient features of the constitutive equations are that they describe history dependent inelastic deformation behaviour of anisotropic metals under three-dimensional stress states in the presence of fast neutron flux. It is shown that the general form of the constitutive relations is consistent with experimental observations made under both unirradiated and irradiated conditions. The utility of the model is demonstrated by examining the analytical results obtained for a segment of tubing undergoing different loading histories in a reactor. (Auth.)

  2. Reactor-core isolation cooling system with dedicated generator

    International Nuclear Information System (INIS)

    Nazareno, E.V.; Dillmann, C.W.

    1992-01-01

    This patent describes a nuclear reactor complex. It comprises a dual-phase nuclear reactor; a main turbine for converting phase-conversion energy stored by vapor into mechanical energy for driving a generator; a main generator for converting the mechanical energy into electricity; a fluid reservoir external to the reactor; a reactor core isolation cooling system with several components at least some of which require electrical power. It also comprises an auxiliary pump for pumping fluid from the reservoir into the reactor pressure vessel; an auxiliary turbine for driving the pump; control means for regulating the rotation rate of the auxiliary turbine; cooling means for cooling the control means; and an auxiliary generator coupled to the auxiliary turbine for providing at least a portion of the electrical power required by the components during a blackout condition

  3. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  4. Upgrading of the Munich reactor with a compact core

    International Nuclear Information System (INIS)

    Boening, K.; Glaeser, W.; Meier, J.; Rau, G.; Roehrmoser, A.; Zhang, L.

    1985-01-01

    An extremely small reactor core has been proposed for the project of substantial modernization of the FRM research reactor at Munich. According to the present status this 'compact core' will be a cylinder with a diameter of about 20 cm and 70 cm high. The new high-density U 3 Si/Al dispersion fuel of about 45% enrichment is contained in 20 concentric fuel plate rings. The compact core is surrounded by a large heavy-water tank which will incorporate the user installations (beam tubes and irradiation channels). However, the primary cooling circuit will contain light water which is not only more economic but also essential for the performance of the small core. An important optimization potential to decrease easily the power density peaks in the core is to reduce further the enrichment in those fuel plate rings where the neutron flux is particularly high. Two-dimensional neutron transport calculations show that such a core, containing about 7.5 kg 235 U, should have an effective multiplication factor of about 1.22 and an unperturbed but realistic maximum thermal neutron flux in the heavy water tank of 7 to 8x10 14 cm -2 .s -1 at 20 MW reactor power. (author)

  5. In-core gamma dosimetry by solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Khan, H.A.

    1980-02-01

    Results are reported of a study undertaken to develop Solid State Nuclear Track Detectors (SSNTD) for the measurement of gamma doses in the megarad region such as those existing in and around a nuclear reactor core. The changes brought about in the track etching parameters and in the ultraviolet and infrared transmittances, have been studied for possible use as gamma dose measuring indices. Effects of various parameters in the core such as neutron flux, beta particles, water, temperature, and gamma ray spectrum have been investigated and found to have only small influence on the proposed gamma dose measuring indices

  6. The solid coolant and prospects of its use in innovative reactors

    International Nuclear Information System (INIS)

    Dmitriev, A.M.; Deniskin, V.P.

    2010-01-01

    The progress of nuclear power demands consideration and development of innovative projects of the reactors having the increased level of safety due to their immanent properties allowing to provide high parameters. One of interesting and perspective offers is the use of a solid substance as a coolant. Use of the solid coolant of a nuclear reactor core has significant advantages among which an opportunity of movement of the coolant in the core under action of gravities and absence of necessity to have superfluous pressure in the jacket, that in turn means small metal consumption of construction, decrease in risk of emergency and its consequences. Cooling of the core with the help of solid substance is possible at performance of the certain conditions connected to features of the solid coolant. The major requirements are: the uniform continuous movement and minimal fluctuation of its density on every site of the core; high mechanical durability and wear resistance of particles; as well as good parameters of heat exchange, i.e. high heat conductivity and thermal capacity of the coolant material at the core operating conditions

  7. Experimental research of reactor core flooding

    International Nuclear Information System (INIS)

    Blaha, V.; Kotrnoch, J.; Krett, V.

    1978-01-01

    The results are presented of experiments performed with the aim of finding the influence of the method of fixing the thermocouples for measuring the distribution of temperature to the wall of fuel pin simulator. This influence was found for the purpose of emergency core flooding. First experimental results on the effect of nitrogen dissolved in the water on the velocity of the cooling wave are given. These experiments were carried out under the following conditions: initial temperature in pin centre 300 to 600 degC, velocity of water at the inlet into the measuring section 3.5 to 20 cm/s, and atmospheric pressure in the model. (author)

  8. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was 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. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  9. Modelling perspectives on radiation chemistry in BWR reactor core

    International Nuclear Information System (INIS)

    Ibe, Eishi

    1991-01-01

    Development of a full-system boiling water reactor core model started in 1982. The model included a two-region reactor core, one with and one without boiling. Key design parameters consider variable dose rates in a three-layer liquid downcomer. Dose rates in the core and downcomer include both generation and recombination reactions of species. Agreement is good between calculations and experimental data of oxygen concentration as a function of hydrogen concentration for different bubble sizes. Oxygen concentration is reduced in the reactor pressure vessel (RPV) by increasing bubble size. The multilayer model follows the oxygen data better than a single-layered model at high concentrations of hydrogen. Key reactions are reduced to five radiolysis reactions and four decomposition reactions for hydrogen peroxide. Calculations by the DOT 3 code showed dose rates from neutrons and gamma rays in various parts of the core. Concentrations of oxygen, hydrogen peroxide, and hydrogen were calculated by the model as a function of time from core inlet. Similar calculations for NWC and HWC were made as a function of height from core inlet both in the boiling channel an the bypass channel. Finally the model was applied to calculate the oxygen plus half the hydrogen peroxide concentrations as a function of hydrogen concentration to compare with data from five plants. Power density distribution with core height was given for an early stage and an end stage of a cycle. Increases of dose rates in the turbine for seven plants were shown as a function of increased hydrogen concentration in the reactor water

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

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

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

  11. Feasibility study of full-reactor gas core demonstration test

    Science.gov (United States)

    Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

    1973-01-01

    Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  13. Nonlinear seismic analysis of a graphite reactor core

    International Nuclear Information System (INIS)

    Laframboise, W.L.; Desmond, T.P.

    1988-01-01

    Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

  14. Solving the uncommon nuclear reactor core neutronics problems

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-01-01

    Calculational procedures have been implemented for solving importance and higher harmonic neutronics problems. Solutions are obtained routinely to support analysis of reactor core performance, treating up to three space coordinates with the multigroup diffusion theory approximation to neutron transport. The techniques used and some of the calculational difficulties are discussed

  15. Use of stainless steel as structural materials in reactor cores

    International Nuclear Information System (INIS)

    Teodoro, C.A.

    1990-01-01

    Austenitic stainless steels are used as structural materials in reactor cores, due to their good mechanical properties at working temperatures and high generalized corrosion resistance in aqueous medium. The objective of this paper is to compare several 300 series austenitic stainless steels related to mechanical properties, localized corrosion resistance (SCC and intergranular) and content of delta ferrite. (author)

  16. Nuclear detectors for in-core power-reactors

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1979-12-01

    Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

  17. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

    Ponzoni Filho, P.; Fernandes, V.B.; Lima Bezerra, J. de; Santos, T.I.C.

    1992-12-01

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

  18. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1990-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

  19. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1989-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

  20. Neutron radiography (NRAD) reactor 64-element core upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately ±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  1. In-core fuel management programs for nuclear power reactors

    International Nuclear Information System (INIS)

    1984-10-01

    In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

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

    Directory of Open Access Journals (Sweden)

    Hao Sun

    2018-03-01

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

  3. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    Martelli, A.; Maresca, G.

    1984-01-01

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  4. Emergency core cooling system in BWR type reactors

    International Nuclear Information System (INIS)

    Takizawa, Yoji

    1981-01-01

    Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

  5. MOX fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Shimada, Hidemitsu; Koyama, Jun-ichi; Aoyama, Motoo

    1998-01-01

    The MOX fuel assembly of the present invention is of a c-lattice type loaded to a BWR type reactor. 74 MOX fuel rods filled with mixed oxides of uranium and plutonium and two water rods disposed to a space equal to that for 7 MOX fuel rods are arranged in 9 x 9 matrix. MOX fuel rods having the lowest enrichment degree are disposed to four corners of the 9 x 9 matrix. The enrichment degree means a ratio of the weight of fission products based on the total weight of fuels. Two MOX fuel rods having the same enrichment degree are arranged in each direction so as to be continuous from the MOX fuel rods at four corners in the direction of the same row and different column and same column and the different row. In addition, among the outermost circumferential portion of the 9 x 9 matrix, MOX fuel rods having a lower enrichment degree next to the MOX fuel rods having the lowest enrichment degree are arranged, each by three to a portion where MOX fuel rods having the lowest enrichment degree are not disposed. (I.N.)

  6. Temperature measurements inside nuclear reactor cores

    International Nuclear Information System (INIS)

    Tarassenko, Serge

    1969-11-01

    Non negligible errors may happen in nuclear reactor temperature measurements using magnesium oxide insulated and stainless steel sheathed micro-wire thermocouples, when these thermometric lines are placed under operational conditions typical of electrical power stations. The present work shows that this error is principally due to electrical hysteresis and polarization phenomena in the insulator subjected to the strong fields generated by common-mode voltages. These phenomena favour the unsymmetrical common-mode current flow and thus lead to the differential-mode voltage generation which is superposing on the thermoelectric hot junction potential. A calculation and an experimental approach make possible the importance of the magnesium oxide insulating characteristics, the hot junction insulation, the choice of the main circuits in the data processing equipment as well as the galvanic isolation performances and the common-mode rejection features of all the measurement circuits. A justification is thereby given for the severe conditions imposed for the acceptance of thermoelectric materials; some particular precautions to be taken are described, as well as the high performance characteristics which have to be taken into account in choosing measurement systems linked to thermometric circuits with sheathed micro-wire thermocouples. (author) [fr

  7. Primary circuit and reactor core T-H characteristics determination of WWER 440 reactors

    International Nuclear Information System (INIS)

    Hermansky, J.; Petenyi, V.; Zavodsky, M.

    2010-01-01

    The WWER-440 nuclear fuel vendor permanently improves the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. During unit refuelling there also could be made some other changes in hydraulic parameters of primary circuit (change of impeller wheels, hydraulic resistance coefficient changes of internal parts of primary circuit, etc.). Therefore it is necessary to determine real coolant flow rate through the reactor during units start-up after their refuelling, and also to have the skilled methodology and computing code for analyzing factors, which affecting the inaccuracy of coolant flow redistribution determination through reactor on flows through separate parts of reactor core in any case of parallel operation of different assembly types. Computing code TH-VCR and CORFLO are used for reactor core characteristics determination for one type of fuel and control assemblies and also in case of parallel operation of different assembly types. The code TH-VCR is able to calculate coolant flow rate for different combinations of three different fuel assembly channel types and three different control assembly channel types. The CORFLO code deals the area of the reactor core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is taken into account. Computing code CORFLO is verified and validated at this time. Paper presents some results from measurements of coolant flow rate through reactors during start-up after unit refuelling and short description of computing code TH-VCR and CORFLO with some calculated results. (Authors)

  8. Hydrogen production with fully integrated fuel cycle gas and vapour core reactors

    International Nuclear Information System (INIS)

    Anghaie, S.; Smith, B.

    2004-01-01

    This paper presents results of a conceptual design study involving gas and vapour core reactors (G/VCR) with a combined scheme to generate hydrogen and power. The hydrogen production schemes include high temperature electrolysis as well as two dominant thermochemical hydrogen production processes. Thermochemical hydrogen production processes considered in this study included the calcium-bromine process and the sulphur-iodine processes. G/VCR systems are externally reflected and moderated nuclear energy systems fuelled by stable uranium compounds in gaseous or vapour phase that are usually operated at temperatures above 1500 K. A gas core reactor with a condensable fuel such as uranium tetrafluoride (UF 4 ) or a mixture of UF 4 and other metallic fluorides (BeF 2 , LiF, KF, etc.) is commonly known as a vapour core reactor (VCR). The single most relevant and unique feature of gas/vapour core reactors is that the functions of fuel and coolant are combined into one. The reactor outlet temperature is not constrained by solid fuel-cladding temperature limits. The maximum fuel/working fluid temperature in G/VCR is only constrained by the reactor vessel material limits, which is far less restrictive than the fuel clad. Therefore, G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Gas and vapour fuel reactors feature very low fuel inventory and fully integrated fuel cycle that provide for exceptional sustainability and safety characteristics. With respect to fuel utilisation, there is no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to the flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from completely burner to breeder is achievable. The continuous recycling of fuel in G/VCR systems allow for complete burning of actinides without removing and reprocessing of the fuel. The only waste products at the back

  9. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1980-01-01

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  10. Computer based core monitoring system for an operating CANDU reactor

    International Nuclear Information System (INIS)

    Yoon, Moon Young; Kwon, O Hwan; Kim, Kyung Hwa; Yeom, Choong Sub

    2004-01-01

    The research was performed to develop a CANDU-6 Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong nuclear power plant unit 1. The CCMS uses Reactor Fueling Simulation Program(RFSP, developed by AECL) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from Digital Control Computer(DCC) for the purpose of producing basic input data. The CCMS has two modules; CCMS server program and CCMS client program. The CCMS server program performs automatic and continuous core calculation and manages overall output controlled by DataBase Management System. The CCMS client program enables users to monitor current and past core status in the predefined GUI(Graphic-User Interface) environment. For the purpose of verifying the effectiveness of CCMS, we compared field-test data with the data used for Wolsong unit 1 operation. In the verification the mean percent differences of both cases were the same(0.008%), which showed that the CCMS could monitor core behaviors well

  11. Core monitoring system for FBR type reactor

    International Nuclear Information System (INIS)

    Azekura, Kazuo.

    1981-01-01

    Purpose: To determine power distribution ON-line after the change of the insertion degree of control rods by the provision of means for calculating power change coefficient at each of the points due to the change in the insertion degree from the specific change of insertion degree and multiplying the same with the newest power distribution determined periodically by the diffusion calculation. Constitution: The monitoring system additionally comprises a calculation device for power change coefficient that calculates the power change coefficient in a fuel assembly adjacent to a control rod based on the data concerning the operation of the control rod, and a provisional power distribution calculation device that executes multiplication between the power distribution calculated in a periodical power distribution calculation device based on the calculation instruction and stored in the core and the power change coefficient from the power change coefficient calculation device and forecasts the provisional power distribution. Then, based on the result of the foregoing calculations, 2-dimensional power distribution, maximum temperature for the cladding tube of the specified fuel assembly, maximum temperature of pellets in the specified fuel assembly, maximum power density and the like are calculated in various display value calculation devices and displayed on a display device. (Horiuchi, T.)

  12. Reference Monte Carlo calculations of Maria reactor core

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.

    2002-01-01

    The reference Monte Carlo calculations of MARIA reactor core have been carried to evaluate accuracy of the calculations at each stage of its neutron-physics analysis using deterministic codes. The elementary cell has been calculated with two main goals; evaluation of effects of simplifications introduced in deterministic lattice spectrum calculations by the WIMS code and evaluation of library data in recently developed WIMS libraries. In particular the beryllium data of those libraries needed evaluation. The whole core calculations mainly the first MARIA critical experiment and the first critical core after the 8-year break in operation. Both cores contained only fresh fuel elements but only in the first critical core the beryllium blocks were not poisoned by Li-6 and He-3. Thus the MCNP k-eff results could be compared with the experiment. The MCNP calculations for the cores with beryllium poisoned suffered the deficiency of uncertainty in the poison concentration, but a comparison of power distribution shows that realistic poison levels have been carried out for the operating reactor MARIA configurations. (author)

  13. Core fusion accidents in nuclear power reactors. Knowledge review

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

  14. Monitoring device for the stability of a reactor core

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To avoid unnecessary limitation on the operation conditions for maintaining the reactor stability. Constitution: The reactor stability is judged by taking notice of the axial power distribution of the reactor and monitoring the same online. Specifically, signals are received from a plurality of local power distribution detectors arranged axially in the reactor core to calculate the axial power distribution in computer. Further, a certain distance L is set from the lower end of the reactor core and the total value S1 for the power distribution in the region below the set value L and the total value S2 for the region above the set value L are determined based on the thus calculated power distribution, to thereby determine the ratio: R = S1/S2 between them. Separately, a certain value r is previously determined based on analysis or experiment such as the result of operation. Then, R and r are compared in a comparator and an alarm is generated, if R >r, with respect to the stability. Since monitoring is made based on the actual index, the applicable range of the operation region can be extended. (Ikeda, J.)

  15. Monte Carlo analysis of Musashi TRIGA mark II reactor core

    International Nuclear Information System (INIS)

    Matsumoto, Tetsuo

    1999-01-01

    The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k eff ) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k eff overestimated the experimental data by about 1.0%Δk/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)

  16. On-site releases of noble gases and iodine in the event of core meltdown in a swimming pool reactor

    International Nuclear Information System (INIS)

    Montaignac, E. de.

    1976-10-01

    Research aimed at defining a standard model accident for swimming pool type reactors, has led to the adoption to the so-called BORAX accident which involves complete meltdown of the reactor core. This type of accident-an accident related to dimensional problems- is useful for calculations concerning reactor components which have to withstand the mechanical forces resulting from the accident. A study of the radiobiological consequences of this type of accident, involving the entire reactor core, required research to determine as accurately as possible how the iodine, noble gases and solid fission products are distributed between the melted core and the site. The joint document in the annexure served as the basis for discussion at the meeting (BEVS/SESR) on 9th March 1973, at which the SESR set the standard parameter values to be used for estimating fission product distributions on the site. (author)

  17. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  18. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

    The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)

  19. Hyper-heuristic applied to nuclear reactor core design

    International Nuclear Information System (INIS)

    Domingos, R P; Platt, G M

    2013-01-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  20. Simulated annealing algorithm for reactor in-core design optimizations

    International Nuclear Information System (INIS)

    Zhong Wenfa; Zhou Quan; Zhong Zhaopeng

    2001-01-01

    A nuclear reactor must be optimized for in core fuel management to make full use of the fuel, to reduce the operation cost and to flatten the power distribution reasonably. The author presents a simulated annealing algorithm. The optimized objective function and the punishment function were provided for optimizing the reactor physics design. The punishment function was used to practice the simulated annealing algorithm. The practical design of the NHR-200 was calculated. The results show that the K eff can be increased by 2.5% and the power distribution can be flattened

  1. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

  2. Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

    International Nuclear Information System (INIS)

    Chow, S.

    1976-01-01

    A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

  3. Hydraulic Profiling of a Parallel Channel Type Reactor Core

    International Nuclear Information System (INIS)

    Seo, Kyong-Won; Hwang, Dae-Hyun; Lee, Chung-Chan

    2006-01-01

    An advanced reactor core which consisted of closed multiple parallel channels was optimized to maximize the thermal margin of the core. The closed multiple parallel channel configurations have different characteristics to the open channels of conventional PWRs. The channels, usually assemblies, are isolated hydraulically from each other and there is no cross flow between channels. The distribution of inlet flow rate between channels is a very important design parameter in the core because distribution of inlet flow is directly proportional to a margin for a certain hydraulic parameter. The thermal hydraulic parameter may be the boiling margin, maximum fuel temperature, and critical heat flux. The inlet flow distribution of the core was optimized for the boiling margins by grouping the inlet orifices by several hydraulic regions. The procedure is called a hydraulic profiling

  4. 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

  5. The application of mechanical desktop in the design of the reactor core structure of China advanced research reactor

    International Nuclear Information System (INIS)

    Lang Ruifeng

    2002-01-01

    The three-dimensional parameterization design method is introduced to the design of reactor core structure for China advanced research reactor. Based on the modeling and dimension variable driving of the main parts as well as the modification of dimension variable, the preliminary design and modification of reactor core is carried out with high design efficiency and quality as well as short periods

  6. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  7. On disruption of reactor core of the Chernobylsk-4 reactor (retrospective analysis of experiments and facts)

    International Nuclear Information System (INIS)

    Platonov, P.A.

    2007-01-01

    Fragments of graphite blocks from the damaged Chernobyl NPP, unit 4 are studied, the results are analyzed. The temperature of the graphite blocks at the moment of accident release from the reactor is evaluated. Results of studying the fragments of fuel channel and fuel dispersion are considered. The fuel heat content at the moment of the explosion is evaluated and some conclusions are made about the character of the reactor core destruction [ru

  8. Reactor core with rod-shaped fuel cells

    International Nuclear Information System (INIS)

    Dworak, A.

    1976-01-01

    The proposal refers to the optimization of the power distribution in a reactor core which is provided with several successive rod-shaped fuel cells. A uniform power output - especially in radial direction - is aimed at. This is achieved by variation of the dwelling periods of the fuel cells, which have, for this purpose, a fuel mixture changing from layer to layer. The fuel cells with the shortest dwelling period are arranged near the coolant inlet side of the reactor core. The dwelling periods of the fuel cells are adapted to the given power distribution. As neighboring cells have equal dwelling periods, the exchange can be performed much easier then with the composition currently known. (UWI) [de

  9. The Core Conversion of the TRIGA Reactor Vienna

    International Nuclear Information System (INIS)

    Villa, M.; Bergmann, R.; Musilek, A.; Sterba, J.H.; Böck, H.; Messick, C.

    2016-01-01

    The TRIGA Reactor Vienna has operated for many years with a mixed core using Al-clad and stainless-steel (SST) clad low enriched uranium (LEU) fuel and a few SST high enriched uranium (HEU) fuel elements. In view of the US spent fuel return program, the average age of these fuel elements and the Austrian position not to store any spent nuclear fuel on its territory, negotiation started in April 2011 with the US Department of Energy (DOE) and the International Atomic Energy Agency (IAEA). The sensitive subject was to return the old TRIGA fuel and to find a solution for a possible continuation of reactor operation for the next decades. As the TRIGA Vienna is the closest nuclear facility to the IAEA headquarters, high interest existed at the IAEA to have an operating research reactor nearby, as historically close cooperation exists between the IAEA and the Atominstitut. Negotiation started before summer 2011 between the involved Austrian ministries, the IAEA and the US DOE leading to the following solution: Austria will return 91 spent fuel elements to the Idaho National Laboratory (INL) while INL offers 77 very low burnt SST clad LEU elements for further reactor operation of the TRIGA reactor Vienna. The titles of these 77 new fuel elements will be transferred to Euratom in accordance with Article 86 of the Euratom-US Treaty. The fuel exchange with the old core returned to the INL, and the new core transferred to Vienna was carried out in one shipment in late 2012 through the ports of Koper/Slovenia and Trieste/Italy. This paper describes the administrative, logistic and technical preparations of the fuel exchange being unique world-wide and first of its kind between Austria and the USA performed successfully in early November 2012. (author)

  10. Shock absorber in combination with a nuclear reactor core structure

    International Nuclear Information System (INIS)

    Housman, J.J.

    1976-01-01

    This invention relates to the provision of shock absorbers for use in blind control rod passages of a nuclear reactor core structure which are not subject to degradation. The shock absorber elements are made of a porous brittle carbonaceous material, a porous brittle ceramic material, or a porous brittle refractory oxide and have a void volume of between 30% and 70% of the total volume of the element for energy absorption by fracturing due to impact loading by a control rod. (UK)

  11. Small ex-core heat pipe thermionic reactor concept (SEHPTR)

    International Nuclear Information System (INIS)

    Jacox, M.G.; Bennett, R.G.; Lundberg, L.B.; Miller, B.G.; Drexler, R.L.

    1991-01-01

    The Idaho National Engineering Laboratory (INEL) has developed an innovative space nuclear power concept with unique features and significant advantages for both Defense and Civilian space missions. The Small Ex-core Heat Pipe Thermionic Reactor (SEHPTR) concept was developed in response to Air Force needs for space nuclear power in the range of 10 to 40 kilowatts. This paper describes the SEHPTR concept and discusses the key technical issues and advantages of such a system

  12. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, T.

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

  13. Fast reactor calculational route for Pu burning core design

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, S. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    This document provides a description of a calculational route, used in the Reactor Physics Research Section for sensitivity studies and initial design optimization calculations for fast reactor cores. The main purpose in producing this document was to provide a description of and user guides to the calculational methods, in English, as an aid to any future user of the calculational route who is (like the author) handicapped by a lack of literacy in Japanese. The document also provides for all users a compilation of information on the various parts of the calculational route, all in a single reference. In using the calculational route (to model Pu burning reactors) the author identified a number of areas where an improvement in the modelling of the standard calculational route was warranted. The document includes comments on and explanations of the modelling assumptions in the various calculations. Practical information on the use of the calculational route and the computer systems is also given. (J.P.N.)

  14. Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

    Science.gov (United States)

    Darmawan, R.

    2018-01-01

    Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

  15. High-temperature process heat reactor with solid coolant and radiant heat exchange

    International Nuclear Information System (INIS)

    Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

    1984-01-01

    The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion

  16. Reactor core materials research and integrated material database establishment

    International Nuclear Information System (INIS)

    Ryu, Woo Seog; Jang, J. S.; Kim, D. W.

    2002-03-01

    Mainly two research areas were covered in this project. One is to establish the integrated database of nuclear materials, and the other is to study the behavior of reactor core materials, which are usually under the most severe condition in the operating plants. During the stage I of the project (for three years since 1999) in- and out of reactor properties of stainless steel, the major structural material for the core structures of PWR (Pressurized Water Reactor), were evaluated and specification of nuclear grade material was established. And the damaged core components from domestic power plants, e.g. orifice of CVCS, support pin of CRGT, etc. were investigated and the causes were revealed. To acquire more resistant materials to the nuclear environments, development of the alternative alloys was also conducted. For the integrated DB establishment, a task force team was set up including director of nuclear materials technology team, and projector leaders and relevant members from each project. The DB is now opened in public through the Internet

  17. Development of an automated core model for nuclear reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input

  18. Conceptual research on reactor core physics for accelerator driven sub-critical reactor

    International Nuclear Information System (INIS)

    Zhao Zhixiang; Ding Dazhao; Liu Guisheng; Fan Sheng; Shen Qingbiao; Zhang Baocheng; Tian Ye

    2000-01-01

    The main properties of reactor core physics are analysed for accelerator driven sub-critical reactor. These properties include the breeding of fission nuclides, the condition of equilibrium, the accumulation of long-lived radioactive wastes, the effect from poison of fission products, as well as the thermal power output and the energy gain for sub-critical reactor. The comparison between thermal and fast system for main properties are carried out. The properties for a thermal-fast coupled system are also analysed

  19. Prevention device for rapid reactor core shutdown in BWR type reactors

    International Nuclear Information System (INIS)

    Koshi, Yuji; Karatsu, Hiroyuki.

    1986-01-01

    Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

  20. Monitoring of core barrel vibrations in WWER type reactor using out-of-reactor ionization chambers

    International Nuclear Information System (INIS)

    Dach, K.

    1982-01-01

    Vibration of the core barrel is least desirable for safe operation of the PWR reactor. These mechanical vibrations are in correlation with the fluctuations of neutron flux density whose time and frequency analysis serves failure diagnosis. The mathematical model is described of the transfer of mechanical vibrations of the core barrel to neutron noise. Other steps are indicated indispensable for the application of the method of neutron noise analysis for in-service diagnostics of nuclear power plants. (Z.M.)

  1. Implementation of new core cooling monitoring system for light water reactors - BCCM (Becker Core Cooling Monitor)

    International Nuclear Information System (INIS)

    Coville, Patrick; Eliasson, Bengt; Stromqvist, Erik; Ward, Olav; Fox, Georges; Ashjian, D. T.

    1998-01-01

    Core cooling monitors are key instruments to protect reactors from large accidents due to loss of coolant. Sensors presented here are based on resistance thermometry. Temperature dependent resistance is powered by relatively high and constant current. Value of this resistance depends on thermal exchange with coolant and when water is no more surrounding the sensors a large increase of temperature is immediately generated. The same instrument can be operated with low current and will measure the local temperature up to 1260 o C in case of loss of coolant accident. Sensors are manufactured with very few components and materials already qualified for long term exposure to boiling or pressurized water reactors environment. Prototypes have been evaluated in a test loop up to 160 bars and in the Barsebaeck-1 reactor. Industrial sensors are now in operation in reactor Oskarshamn 2. (author)

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

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

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

  3. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Directory of Open Access Journals (Sweden)

    Krása Antonín

    2017-01-01

    Full Text Available VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector. Discrepancies between experiments and Monte Carlo calculations (MCNP5 of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2 are presented.

  4. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Science.gov (United States)

    Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

    2017-09-01

    VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

  5. Thermal hydraulics model for Sandia's annular core research reactor

    International Nuclear Information System (INIS)

    Rao, Dasari V.; El-Genk, Mohamed S.; Rubio, Reuben A.; Bryson, James W.; Foushee, Fabian C.

    1988-01-01

    A thermal hydraulics model was developed for the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The coupled mass, momentum and energy equations for the core were solved simultaneously using an explicit forward marching numerical technique. The model predictions of the temperature rise across the central channel of the ACRR core were within ± 10 percent agreement with the in-core temperature measurements. The model was then used to estimate the coolant mass flow rate and the axial distribution of the cladding surface temperature in the central and average channels as functions of the operating power and the water inlet subcooling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the ACRR at power levels in excess of 0.5 MW. However, the high heat transfer coefficient due to subcooled boiling causes the cladding temperature along most of the active fuel rod region to be quite uniform and to increase very little with the reactor power. (author)

  6. In-core fuel management for nuclear reactor

    International Nuclear Information System (INIS)

    Ross, M.F.; Visner, S.

    1986-01-01

    This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

  7. Neutronic design of a traveling wave reactor core

    International Nuclear Information System (INIS)

    Lopez S, R. C.; Francois L, J. L.

    2010-10-01

    The traveling wave reactor is an innovative kind of fast breeder reactor, capable of operate for decades without refueling and whose operation requires only a small amount of enriched fuel for the ignition. Also, one of its advantages is its versatility; it can be designed as small modules of about 100 M We or large scale units of 1000 M We. In this paper the behaviour of the traveling wave reactor core is studied in order to determine whether the traveling breeding/burning wave moves (as theoretically predicted) or not. To achieve this, we consider a two pieces cylinder, the first one, the ignition zone, containing highly enriched fuel and the second, the breeding zone, which is the larger, containing natural or depleted uranium or thorium. We consider that both zones are homogeneous mixtures of fuel, sodium as coolant and iron as structural material. We also include a reflector material outside the cylinder to reduce the neutron leakages. Simulations were run with MCNPX version 2.6 code. We observed that the wave does move as time passes as predicted by theory, and reactor remains supercritical in the time we have simulated (3000 days). Also, we found that thorium does not perform as well as uranium for breeding in this type of reactor. Further test with different reflectors are planned for both U-Pu and Th-U fuel cycles. (Author)

  8. Real-time advanced nuclear reactor core model

    International Nuclear Information System (INIS)

    Koclas, J.; Friedman, F.; Paquette, C.; Vivier, P.

    1990-01-01

    The paper describes a multi-nodal advanced nuclear reactor core model. The model is based on application of modern equivalence theory to the solution of neutron diffusion equation in real time employing the finite differences method. The use of equivalence theory allows the application of the finite differences method to cores divided into hundreds of nodes, as opposed to the much finer divisions (in the order of ten thousands of nodes) where the unmodified method is currently applied. As a result the model can be used for modelling of the core kinetics for real time full scope training simulators. Results of benchmarks, validate the basic assumptions of the model and its applicability to real-time simulation. (orig./HP)

  9. Pebble bed reactor with one-zone core

    International Nuclear Information System (INIS)

    Mueller-Frank, U.; Lohnert, G.

    1977-01-01

    The claim deals with measures to differentiate the flow rate and to remove spherical fuel elements in the core of a pebble bed reactor. Hence the vertical rate of the fuel elements in the border region is for example twice as much as in the centre. A central funnel-shaped outlet on the floor of the core container over which a conical body is placed with its peak pointing upwards, or also the forming of several outlets can be used to adjust to a certain exit rate for the fuel elements. The main target of the invention is a radially extensively constant coolant outlet temperature at the outlet of the core which determines the effectiveness of the connected heat exchanger and thus contributes to economy. (UA) [de

  10. Thermal radiation in gas core nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J.

    1994-01-01

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs

  11. Solid charged-core model of ball lightning

    Science.gov (United States)

    Muldrew, D. B.

    2010-01-01

    In this study, ball lightning (BL) is assumed to have a solid, positively-charged core. According to this underlying assumption, the core is surrounded by a thin electron layer with a charge nearly equal in magnitude to that of the core. A vacuum exists between the core and the electron layer containing an intense electromagnetic (EM) field which is reflected and guided by the electron layer. The microwave EM field applies a ponderomotive force (radiation pressure) to the electrons preventing them from falling into the core. The energetic electrons ionize the air next to the electron layer forming a neutral plasma layer. The electric-field distributions and their associated frequencies in the ball are determined by applying boundary conditions to a differential equation given by Stratton (1941). It is then shown that the electron and plasma layers are sufficiently thick and dense to completely trap and guide the EM field. This model of BL is exceptional in that it can explain all or nearly all of the peculiar characteristics of BL. The ES energy associated with the core charge can be extremely large which can explain the observations that occasionally BL contains enormous energy. The mass of the core prevents the BL from rising like a helium-filled balloon - a problem with most plasma and burning-gas models. The positively charged core keeps the negatively charged electron layer from diffusing away, i.e. it holds the ball together; other models do not have a mechanism to do this. The high electrical charges on the core and in the electron layer explains why some people have been electrocuted by BL. Experiments indicate that BL radiates microwaves upon exploding and this is consistent with the model. The fact that this novel model of BL can explain these and other observations is strong evidence that the model should be taken seriously.

  12. Aspects of cell calculations in deterministic reactor core analysis

    International Nuclear Information System (INIS)

    Varvayanni, M.; Savva, P.; Catsaros, N.

    2011-01-01

    Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for

  13. GCRA review and appraisal of HTGR reactor-core-design program

    International Nuclear Information System (INIS)

    1980-09-01

    The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation

  14. Theoretical studies on core-level spectra of solids

    International Nuclear Information System (INIS)

    Kotani, Akio

    1995-01-01

    I present a review on theoretical studies of core-level spectra (CLS) in solids. In CLS, the dynamical response of outer electrons to a core hole is reflected through the screening of core hole potential. Impurity Anderson model (IAM) or cluster model is successfully applied to the analysis of X-ray photoemission spectra (XPS) and X-ray absorption spectra (XAS) in f and d electron systems, where the f and d electron states are hybridized with the other valence or conduction electron states. The effect of the core-hole potential in the final state of XPS and XAS plays an important role, as well as the solid state hybridization and intra-atomic multiplet coupling effects. As typical examples, the calculated results for XPS of rare-earth compounds and transition metal compounds are shown, and some discussions are given. As a subject of remarkable progress with high brightness synchrotron radiation sources, I discuss some theoretical aspects of X-ray emission spectra (XES) and their resonant enhancement at the X-ray absorption threshold. Some experimental data and their theoretical analysis are also given. (author)

  15. Neutronic design of mixed oxide-silicide cores for the core conversion of rsg-gas reactor

    International Nuclear Information System (INIS)

    Sembiring, Tagor Malem; Tukiran; Pinem surian; Febrianto

    2001-01-01

    The core conversion of rsg-gas reactor from an all-oxide (U 3 O 8 -Al) core, through a series of mixed oxide-silicide core, to an all-silicide (U 3 Si 2 -Al) core for the same meat density of 2.96 g U/cc is in progress. The conversion is first step of the step-wise conversion and will be followed by the second step that is the core conversion from low meat density of silicide core, through a series of mixed lower-higher density of silicide core, to an all-higher meat density of 3.55 g/cc core. Therefore, the objectives of this work is to design the mixed cores on the neutronic performance to achieve safety a first full-silicide core for the reactor with the low uranium meat density of 2.96gU/cc. The neutronic design of the mixed cores was performed by means of Batan-EQUIL-2D and Batan-3DIFF computer codes for 2 and 3 dimension diffusion calculation, respectively. The result shows that all mixed oxide-silicide cores will be feasible to achieve safety a fist full-silicide core. The core performs the same neutronic core parameters as those of the equilibrium silicide core. Therefore, the reactor availability and utilization during the core conversion is not changed

  16. Emergency core cooling system for LMFBR type reactors

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Fukutomi, Shigeki.

    1980-01-01

    Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

  17. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  18. Solid methane cold moderator for the IBR-2 reactor

    International Nuclear Information System (INIS)

    Beliakov, A. A.; Tretiakov, I. T.; Shabalin, E. P.; Golikov, V. V.; Luschivkov, V I.

    1997-09-01

    The paper describes the research and design work carried out since 1986 at the Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear Research in Dubna to create a cryogenic moderator for the IBR-22 reactor using solid methane as a moderating substance.

  19. Core design study on reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Hiroshi, Akie; Yoshihiro, Nakano; Toshihisa, Shirakawa; Tsutomu, Okubo; Takamichi, Iwamura

    2002-01-01

    The conceptual core design study of reduced-moderation water reactors (RMWRs) with tight-pitched MOX-fuelled lattice has been carried out at JAERI. Several different RMWR core concepts based on both BWR and PWR have been proposed. All the core concepts meet with the aim to achieve both a conversion ratio of 1.0 or larger and negative void reactivity coefficient. As one of these RMWR concepts, the ABWR compatible core is also proposed. Although the conversion ratio of this core is 1.0 and the void coefficient is negative, the discharge burn-up of the fuel was about 25 GWd/t. By adopting a triangular fuel pin lattice for the reduction of moderator volume fraction and modifying axial Pu enrichment distribution, it was aimed to extend the discharge burn-up of ABWR compatible type RMWR. By using a triangular fuel lattice of smaller moderator volume fraction, discharge burn-up of 40 GWd/t seems achievable, keeping the high conversion ratio and the negative void coefficient. (authors)

  20. Seismic responses of a pool-type fast reactor with different core support designs

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Seidensticker, R.W.

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

  1. Reactor Core Internals Replacement of Ikata Units 1 and 2

    International Nuclear Information System (INIS)

    Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

    2012-01-01

    This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

  2. Global physical and numerical stability of a nuclear reactor core

    International Nuclear Information System (INIS)

    Morales-Sandoval, Jaime; Hernandez-Solis, Augusto

    2005-01-01

    Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented

  3. Device for measuring flow rate in a nuclear reactor core

    International Nuclear Information System (INIS)

    Hamano, Jiro.

    1980-01-01

    Purpose: To always calculate core flow rate automatically and accurately in BWR type nuclear power plants. Constitution: Jet pumps are provided to the recycling pump and to the inside of the pressure vessel of a nuclear reactor. The jet pumps comprise a plurality of calibrated jet pumps for forcively convecting the coolants and a plurality of not calibrated jet pumps in order to cool the heat generated in the reactor core. The difference in the pressures between the upper and the lower portions in both of the jet pumps is measured by difference pressure transducers. Further, a thermo-sensitive element is provided to measure the temperature of recycling water at the inlet of the recycling pump. The output signal from the difference pressure transducer is inputted to a process computer, calculated periodically based on predetermined calculation equations, compensated for the temperature by a recycling water temperature signal and outputted as a core flow rate signal to a recoder. The signal is also used for the power distribution calculation in the process computer and the minimum limit power ratio as the thermal limit value for the fuels is outputted. (Furukawa, Y.)

  4. Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhujie [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Bartels, David [Univ. of Notre Dame, IN (United States)

    2015-04-02

    This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.

  5. Reactor vessel and core two-phase flow ultrasonic densitometer

    International Nuclear Information System (INIS)

    Arave, A.E.

    1979-01-01

    A local ultrasonic density (LUD) detector has been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) for the Loss-of-Fluid Test (LOFT) reactor vessel and core two-phase flow density measurements. The principle of operating the sensor is the change in propagation time of a torsional ultrasonic wave in a metal transmission line as a function of the density of the surrounding media. A theoretical physics model is presented which represents the total propagation time as a function of the sensor modulus of elasticity and polar moment of inertia. Separate effects tests and two-phase flow tests have been conducted to characterize the detector. Tests show the detector can perform in a 343 0 C pressurized water reactor environment and measure the average density of the media surrounding the sensor

  6. Examination of core components removed from CANDU reactors

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Coleman, C.E.; Rodgers, D.K.; Davies, P.H.; Chow, C.K.; Griffiths, M.

    1988-11-01

    Components in the core of a nuclear reactor degrade because the environment is severe. For example, in CANDU reactors the pressure tubes must contend with the effects of hot pressurised water and damage by a flux of fast neutrons. To evaluate any deterioration of components and determine the cause of the occasional failure, we have developed a wide range of remote-handling techniques to examine radioactive materials. As well as pressure tubes, we have examined calandria tubes, garter springs, end fittings, liquid-zone control units and flux detectors. The results from these examinations have produced solutions to problems and continually provide information to help understand the processes that may limit the lifetime of a component

  7. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

    Carew, J.F.; Neogy, P.

    1983-01-01

    An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

  8. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

    1999-01-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  9. Solid oxide fuel cell having a monolithic core

    International Nuclear Information System (INIS)

    Ackerman, J.P.; Young, J.E.

    1984-01-01

    A solid oxide fuel cell for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween, and each interconnect wall consists of thin layers of the cathode and anode materials sandwiching a thin layer of interconnect material therebetween. The electrolyte walls are arranged and backfolded between adjacent interconnect walls operable to define a plurality of core passageways alternately arranged where the inside faces thereof have only the anode material or only the cathode material exposed. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageway; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte and interconnect materials is of the order of 0.002-0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002-0.05 cm thick

  10. A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Rui, E-mail: rhu@anl.gov; Yu, Yiqi

    2016-11-15

    Highlights: • Developed a computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The accuracy and efficiency of the method is confirmed with a 7-assembly test problem. • The effects of different spatial discretization schemes are investigated and compared to the RANS-based CFD simulations. - Abstract: For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. Additionally, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

  11. Tritium dynamics in fusion reactor solid breeder

    International Nuclear Information System (INIS)

    Violante, V.

    1986-01-01

    In the field of the NET research progrm, the chemical and diffusive processes involved in solid ceramic breeder materials have been analysed. A mathematical model describing the phenomena has been developed to obtain a quantitative evaluation for a first design approach. The data obtained by means of the above mentioned model are in good agreement with the data obtained by other research groups working in Europe and in United States. The computer codes BLANKET2, MC2, FWBC, have been developed to simulate the phenomena

  12. A supercomputing application for reactors core design and optimization

    International Nuclear Information System (INIS)

    Hourcade, Edouard; Gaudier, Fabrice; Arnaud, Gilles; Funtowiez, David; Ammar, Karim

    2010-01-01

    Advanced nuclear reactor designs are often intuition-driven processes where designers first develop or use simplified simulation tools for each physical phenomenon involved. Through the project development, complexity in each discipline increases and implementation of chaining/coupling capabilities adapted to supercomputing optimization process are often postponed to a further step so that task gets increasingly challenging. In the context of renewal in reactor designs, project of first realization are often run in parallel with advanced design although very dependant on final options. As a consequence, the development of tools to globally assess/optimize reactor core features, with the on-going design methods accuracy, is needed. This should be possible within reasonable simulation time and without advanced computer skills needed at project management scale. Also, these tools should be ready to easily cope with modeling progresses in each discipline through project life-time. An early stage development of multi-physics package adapted to supercomputing is presented. The URANIE platform, developed at CEA and based on the Data Analysis Framework ROOT, is very well adapted to this approach. It allows diversified sampling techniques (SRS, LHS, qMC), fitting tools (neuronal networks...) and optimization techniques (genetic algorithm). Also data-base management and visualization are made very easy. In this paper, we'll present the various implementing steps of this core physics tool where neutronics, thermo-hydraulics, and fuel mechanics codes are run simultaneously. A relevant example of optimization of nuclear reactor safety characteristics will be presented. Also, flexibility of URANIE tool will be illustrated with the presentation of several approaches to improve Pareto front quality. (author)

  13. Core disruptive accident analysis in prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Kannan, S.E.; Singh, Om Pal; Chetal, S.C.; Bhoje, S.B.

    2002-01-01

    Liquid metal cooled fast breeder reactors, in particular, pool type have many inherent and engineered safety features and hence a core disruptive accident (CDA) involving melt down of the whole core is a very low probable event ( -6 /ry). The important mechanical consequences such as straining of the main vessel including top shield, structural integrity of safety grade decay heat exchangers (DHX) and intermediate heat exchangers (IHX) sodium release to reactor containment building (RCB) through the penetrations in the top shield, sodium fire and consequent temperature and pressure rise in RCB are theoretically analysed using computer codes. Through the analyses with these codes, it is demonstrated that an energetic CDA capability to the maximum 100 MJ mechanical energy in PFBR can be well contained in the primary containment. The sodium release to RCB is 350 kg and pressure rise in RCB is ∼10 kPa. In order to raise the confidence on the theoretical predictions, very systematic experimental program has been carried out. Totally 67 tests were conducted. This experimental study indicated that the primary containment is integral. The main vessel can withstand the energy release of ∼1200 MJ. The structural integrity of IHX and DHX is assured up to 200 MJ. The transient force transmitted to reactor vault is negligible. The average water leak measured under simulated tests for 122 MJ work potential is about 1.8 kg and the maximum leak is 2.41 kg. Extrapolation of the measured maximum leak based on simulation principles yields ∼ 233 kg of sodium leak in the reactor. Based on the above-mentioned theoretical and experimental investigations, the design pressure of 20 kPa is used for PFBR

  14. Nuclear power reactor core melt accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

    2015-11-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt

  15. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  16. Some experimental justifications of constructions of nuclear reactors with the use of solid coolant

    International Nuclear Information System (INIS)

    Deniskin, V.; Nalivaev, V.; Fedik, I.; Vishnevski, U.; Dmitriev, A.

    2003-01-01

    Full text: The work that has been conducted so far justifies a possibility of constructing a reactor with a non-traditional coolant to develop radically new reactors and their cycles with perfect architecture. A solid coolant, for example, the carbon-based one, allows to design the primary circuit of nuclear reactor without excess pressure. Such coolant withstands temperatures up to ∼4000 deg. K without a collapse. The analysis of theory and experiments produced requirements to be met by a solid coolant used in the primary circuit of nuclear reactor. One of the most important requirements is the arrangements for a continuous and homogeneous gravity flow of the coolant through all core sections taking into account the dust caused by wear and some amount of fractured particles. Therefore, the idea is that the mass of particles should resemble a liquid to a certain extend. The particles should be sphere like with average diameter from 0.5 to 2.0 mm and nonsphericity rate not more than 10%. 'Angle of repose' of particles to the horizon can be utilised as a validity criterion of particles which should not exceed 25 deg. The heat transfer coefficient should be increased up to the practical maximum value. In 1996 - 1997 the system of experimental facilities were built in the Scientific and Research Institute 'Luch' to prove the possibility to reliably cool a nuclear reactor with a flow of solid particles and to obtain a minimum set of data for the conceptual design of such reactor with solid coolant. The facility allows the research of the flow stability, heat mass transfer in the core, lifetime wearing of particles of the solid coolant. In 1994-1999 5 batches of particles of different size were fabricated in accordance to different technologies. Four batches were graphite-based and one was aluminium oxide-based (Al 2 O 3 ). The purpose was to verify how the heat transfer coefficient was changing as the particle size varied. The average diameter of graphite particles

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  18. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  19. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  20. Study of the seismic behaviour of the fast reactor cores

    International Nuclear Information System (INIS)

    Cerqueira, E.

    1998-01-01

    This work studies the seismic behaviour of fast neutrons reactor cores. It consists in analyzing the tests made on the models Rapsodie and Symphony by using the calculation code Castem 2000. Te difficulty is in the description of connections of the system and the effects of the fluid (calculation in water). The results for the programme Rapsodie are near the experimental results. For the programme Symphony, the calculations in air have allowed to represent the behaviour of fuel assemblies in a satisfying way. It is still to analyze the tests Symphony in water. (N.C.)

  1. Nuclear reactor ex-core startup neutron detector

    International Nuclear Information System (INIS)

    Wyvill, J.R.

    1980-01-01

    A sensitive ex-core neutron detector is needed to monitor the power level of reactors during startup. The neutron detector of this invention has a photomultiplier with window resistant to radiation darkening at the input end and an electrical connector at the output end. The photomultiplier receives light signals from a neutron-responsive scintillator medium, typically a cerium-doped lithium silicate glass, that responds to neutrons after they have been thermalized by a silicone resin moderator. Enclosing and shielding the photmultiplier, the scintillator medium and the moderator is a combined lead and borated silicone resin housing

  2. Reactor physics innovations of the advanced CANDU reactor core: adaptable and efficient

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Hopwood, J.M.; Bonechi, M.

    2003-01-01

    The Advanced CANDU Reactor (ACR) is designed to have a benign, operator-friendly core physics characteristic, including a slightly negative coolant-void reactivity and a moderately negative power coefficient. The discharge fuel burnup is about three times that of natural uranium fuel in current CANDU reactors. Key features of the reactor physics innovations in the ACR core include the use of H 2 O coolant, slightly enriched uranium (SEU) fuel, and D 2 O moderator in a reduced lattice pitch. These innovations result in substantial improvements in economics, as well as significant enhancements in reactor performance and waste reduction over the current reactor design. The ACR can be readily adapted to different power outputs by increasing or decreasing the number of fuel channels, while maintaining identical fuel and fuel-channel characteristics. The flexibility provided by on-power refuelling and simple fuel bundle design enables the ACR to easily adapt to the use of plutonium and thorium fuel cycles. No major modifications to the basic ACR design are required because the benign neutronic characteristics of the SEU fuel cycle are also inherent in these advanced fuel cycles. (author)

  3. Status of the Munich compact core reactor project

    International Nuclear Information System (INIS)

    Boening, K.; Glaeser, W.; Meier, J.; Rau, G.; Roehrmoser, A.; Steichele, E.

    1991-01-01

    A new national, high performance research reactor is being planned in Munich which shall provide an unperturbed thermal neutron flux of about 8.10 14 cm -2 s -1 at only 20 MW power. This object can be achieved with the concept of a small, light water cooled 'compact core' situated in the center of a large heavy water moderator tank. In the first part of this paper an overview is given on the status of the necessary fuel enrichment is discussed. In becomes clear that the potential of the compact core concept can only be developed without compromises when using highly enriched uranium in combination with the new high density fuel. (orig.)

  4. Computer code for simulating pressurized water reactor core

    International Nuclear Information System (INIS)

    Serrano, A.M.B.

    1978-01-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)

  5. Code for the core simulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1978-08-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numericaly. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistence added to the film coeficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (Author) [pt

  6. Combustion of metal agglomerates in a solid rocket core flow

    Science.gov (United States)

    Maggi, Filippo; Dossi, Stefano; DeLuca, Luigi T.

    2013-12-01

    The need for access to space may require the use of solid propellants. High thrust and density are appealing features for different applications, spanning from boosting phase to other service applications (separation, de-orbiting, orbit insertion). Aluminum is widely used as a fuel in composite solid rocket motors because metal oxidation increases enthalpy release in combustion chamber and grants higher specific impulse. Combustion process of metal particles is complex and involves aggregation, agglomeration and evolution of reacting particulate inside the core flow of the rocket. It is always stated that residence time should be enough in order to grant complete metal oxidation but agglomerate initial size, rocket grain geometry, burning rate, and other factors have to be reconsidered. New space missions may not require large rocket systems and metal combustion efficiency becomes potentially a key issue to understand whether solid propulsion embodies a viable solution or liquid/hybrid systems are better. A simple model for metal combustion is set up in this paper. Metal particles are represented as single drops trailed by the core flow and reacted according to Beckstead's model. The fluid dynamics is inviscid, incompressible, 1D. The paper presents parametric computations on ideal single-size particles as well as on experimental agglomerate populations as a function of operating rocket conditions and geometries.

  7. Solid-state track recorder neutron dosimetry in light water reactor pressure vessel surveillance mockups

    International Nuclear Information System (INIS)

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

    1984-09-01

    Solid-State Track Recorder (SSTR) measurements of neutron-induced fission rates have been made in several pressure vessel mockup facilities as part of the US Nuclear Regulatory Commission's (NRC) Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP). The results of extensive physics-dosimetry measurements made at the Pool Critical Assembly (PCA) at Oak Ridge National Laboratory (ORNL) in Oak Ridge, TN are summarized. Included are 235 U, 238 U, 237 Np and 232 Th fission rates in the PCA 12/13, 8/7, and 4/12 SSC configurations. Additional low power measurements have been made in an engineering mockup at the VENUS critical assembly at CEN-SCK, Mol, Belgium. 237 Np and 238 U fission rates were made at selected locations in the VENUS mockup, which models the in-core and near-core regions of a pressurized water reactor (PWR). Absolute core power measurements were made at VENUS by exposing solid-state track recorders (SSTRs) to polished fuel pellets within in-core fuel pins. 8 references, 4 figures, 10 tables

  8. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  9. Core optimization studies for a small heating reactor

    International Nuclear Information System (INIS)

    Galperin, A.

    1986-11-01

    Small heating reactor cores are characterized by a high contribution of the leakage to the neutron balance and by a large power density variation in the axial direction. A limited number of positions is available for the control rods, which are necessary to satisfy overall reactivity requirements subject to a safety related constraint on the maximum worth of each rod. Design approaches aimed to improve safety and fuel utilization performance of the core include separation of the cooling and moderating functions of the water with the core in order to reduce hot-to-cold reactivity shift and judicious application of the axial Gd zoning aimed to improve the discharge burnup distribution. Several design options are analyzed indicating a satisfactory solution of the axial burnup distribution problem. The feasibility of the control rod system including zircaloy, stainless steel, natural boron and possibly enriched boron rods is demonstrated. A preliminary analysis indicates directions for further improvements of the core performance by an additional reduction of the hot-to-cold reactivity shift and by a reduction of the depletion reactivity swing adopting a higher gadolinium concentration in the fuel or a two-batch fuel management scheme. (author)

  10. Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

    International Nuclear Information System (INIS)

    S, Tukiran; MS, Tagor; P, Surian

    2003-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

  11. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  12. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  13. Heat transfer in reactor cavity during core-concrete interaction

    International Nuclear Information System (INIS)

    Adroguer, B.; Cenerino, G.

    1989-08-01

    In the unlikely event of a severe accident in a nuclear power plant, the core may melt through the vessel and slump into the concrete reactor cavity. The hot mixture of the core material called corium interacts thermally with the concrete basemat. The WECHSL code, developed at K.f.K. Karlsruhe in Germany is used at the Protection and Nuclear Safety Institute (I.P.S.N.) of CEA to compute this molten corium concrete interaction (MCCI). Some uncertainties remain in the partition of heat from the corium between the basemat and the upper surrounding structures in the cavity where the thermal conditions are not computer. The CALTHER code, under development to perform a more mechanistic evaluation of the upward heat flux has been linked to WECHSL-MOD2 code. This new version enables the modelling of the feedback effects from the conditions in the cavity to the MCCI and the computation of the fraction of upward flux directly added to the cavity atmosphere. The present status is given in the paper. Preliminary calculations of the reactor case for silicate and limestone common sand (L.C.S.) concretes are presented. Significant effects are found on concrete erosion, gases release and temperature of the upper part of corium, particularly for L.C.S. concrete

  14. Determination of temperature distributions in fast reactor core coolants

    International Nuclear Information System (INIS)

    Tillman, M.

    1975-04-01

    An analytical method of determination of a temperature distribution in the coolant medium in a fuel assembly of a liquid-metal-fast-breeder-reactor (LMFBR) is presented. The temperature field obtained is applied for a constant velocity (slug flow) fluid flowing, parallel to the fuel pins of a square and hexagonal array assembly. The coolant subchannels contain irregular boundaries. The geometry of the channel due to the rod adjacent to the wall (edge rod) differs from the geometry of the other channels. The governing energy equation is solved analytically, assuming series solutions for the Poisson and diffusion equations, and the total solution is superposed by the two. The boundary conditions are specified by symmetry considerations, assembly wall insulation and a continuity of the temperature field and heat fluxes. The initial condition is arbitrary. The method satisfies the boundary conditions on the irregular boundaries and the initial condition by a least squares technique. Computed results are presented for various geometrical forms, with ratio of rod pitch-to-diameter typical for LMFBR cores. These results are applicable for various fast-reactors, and thus the influence of the transient solution (which solves the diffusion equation) on the total depends on the core parameters. (author)

  15. A neural network to predict reactor core behaviors

    International Nuclear Information System (INIS)

    Juan Jose Ortiz-Servin; Jose Alejandro Castillo; Pelta, David A.

    2014-01-01

    The global fuel management problem in BWRs (Boiling Water Reactors) can be understood as a very complex optimization problem, where the variables represent design decisions and the quality assessment of each solution is done through a complex and computational expensive simulation. This last aspect is the major impediment to perform an extensive exploration of the design space, mainly due to the time lost evaluating non promising solutions. In this work, we show how we can train a Multi-Layer Perceptron (MLP) to predict the reactor behavior for a given configuration. The trained MLP is able to evaluate the configurations immediately, thus allowing performing an exhaustive evaluation of the possible configurations derived from a stock of fuel lattices, fuel reload patterns and control rods patterns. For our particular problem, the number of configurations is approximately 7.7 x 10 10 ; the evaluation with the core simulator would need above 200 years, while only 100 hours were required with our approach to discern between bad and good configurations. The later were then evaluated by the simulator and we confirm the MLP usefulness. The good core configurations reached the energy requirements, satisfied the safety parameter constrains and they could reduce uranium enrichment costs. (authors)

  16. 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

  17. Machine learning of the reactor core loading pattern critical parameters

    International Nuclear Information System (INIS)

    Trontl, K.; Pevec, D.; Smuc, T.

    2007-01-01

    The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employed a recently introduced machine learning technique, Support Vector Regression (SVR), which has a strong theoretical background in statistical learning theory. Superior empirical performance of the method has been reported on difficult regression problems in different fields of science and technology. SVR is a data driven, kernel based, nonlinear modelling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modelling. The starting set of experimental data for training and testing of the machine learning algorithm was obtained using a two-dimensional diffusion theory reactor physics computer code. We illustrate the performance of the solution and discuss its applicability, i.e., complexity, speed and accuracy, with a projection to a more realistic scenario involving machine learning from the results of more accurate and time consuming three-dimensional core modelling code. (author)

  18. Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

    International Nuclear Information System (INIS)

    Broc, Daniel

    2001-01-01

    As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

  19. Thermohydraulics of emergency core cooling in light water reactors

    International Nuclear Information System (INIS)

    1989-10-01

    This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided

  20. Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores

    International Nuclear Information System (INIS)

    El Ganaoui, K.

    2006-09-01

    In the context of homogenization theory we treat some heat transfer problems involving unusual (according to the homogenization) boundary conditions. These problems are defined in a solid periodic perforated domain where two scales (macroscopic and microscopic) are to be taken into account and describe heat transfer by conduction in the solid and by radiation on the wall of each hole. Two kinds of radiation are considered: radiation in an infinite medium (non-linear problem) and radiation in cavity with grey-diffuse walls (non-linear and non-local problem). The derived homogenized models are conduction problems with an effective conductivity which depend on the considered radiation. Thus we introduce a framework (homogenization and validation) based on mathematical justification using the two-scale convergence method and numerical validation by simulations using the computer code CAST3M. This study, performed for gas cooled reactors cores, can be extended to other perforated domains involving the considered heat transfer phenomena. (author)

  1. Heating analysis of cobalt adjusters in reactor core

    International Nuclear Information System (INIS)

    Mei Qiliang; Li Kang; Fu Yaru

    2011-01-01

    In order to produce 60 Co source for industry and medicine applications in CANDU-6 reactor, the stainless steel adjusters were replaced with the cobalt adjusters. The cobalt rod will generate the heat when it is irradiated by neutron and γ ray. In addition, 59 Co will be activated and become 60 Co, the ray released due to 60 Co decay will be absorbed by adjusters, and then the adjusters will also generate the heat. So the heating rate of adjusters to be changed during normal operation must be studied, which will be provided as the input data for analyzing the temperature field of cobalt adjusters and the relative heat load of moderator. MCNP code was used to simulate whole core geometric configuration in detail, including reactor fuel, control rod, adjuster, coolant and moderator, and to analyze the heating rate of the stainless steel adjusters and the cobalt adjusters. The maximum heating rate of different cobalt adjuster based on above results will be provided for the steady thermal hydraulic and accident analysis, and make sure that the reactor is safe on the thermal hydraulic. (authors)

  2. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes.

  3. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    International Nuclear Information System (INIS)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun

    2016-01-01

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes

  4. Real time thermal hydraulic model for high temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

    2013-01-01

    A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)

  5. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  6. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  7. An integration scheme for stiff solid-gas reactor models

    Directory of Open Access Journals (Sweden)

    Bjarne A. Foss

    2001-04-01

    Full Text Available Many dynamic models encounter numerical integration problems because of a large span in the dynamic modes. In this paper we develop a numerical integration scheme for systems that include a gas phase, and solid and liquid phases, such as a gas-solid reactor. The method is based on neglecting fast dynamic modes and exploiting the structure of the algebraic equations. The integration method is suitable for a large class of industrially relevant systems. The methodology has proven remarkably efficient. It has in practice performed excellent and been a key factor for the success of the industrial simulator for electrochemical furnaces for ferro-alloy production.

  8. Determination of the NPP Krsko reactor core safety limits using the COBRA-III-C code

    International Nuclear Information System (INIS)

    Lajtman, S.; Feretic, D.; Debrecin, N.

    1989-01-01

    This paper presents the NPP Krsko reactor core safety limits determined by the COBRA-III-C code, along with the methodology used. The reactor core safety limits determination is a part of reactor protection limits procedure. The results obtained were compared to safety limits presented in NPP Krsko FSAR. The COBRA-III-C NPP Krsko design core steady state thermal hydraulics calculation, used as the basis for the safety limits calculation, is presented as well. (author)

  9. Static analysis of material testing reactor cores:critical core calculations

    International Nuclear Information System (INIS)

    Nawaz, A. A.; Khan, R. F. H.; Ahmad, N.

    1999-01-01

    A methodology has been described to study the effect of number of fuel plates per fuel element on critical cores of Material Testing Reactors (MTR). When the number of fuel plates are varied in a fuel element by keeping the fuel loading per fuel element constant, the fuel density in the fuel plates varies. Due to this variation, the water channel width needs to be recalculated. For a given number of fuel plates, water channel width was determined by optimizing k i nfinity using a transport theory lattice code WIMS-D/4. The dimensions of fuel element and control fuel element were determined using this optimized water channel width. For the calculated dimensions, the critical cores were determined for the given number of fuel plates per fuel element by using three dimensional diffusion theory code CITATION. The optimization of water channel width gives rise to a channel width of 2.1 mm when the number of fuel plates is 23 with 290 g ''2''3''5U fuel loading which is the same as in the case of Pakistan Reactor-1 (PARR-1). Although the decrease in number of fuel element results in an increase in optimal water channel width but the thickness of standard fuel element (SFE) and control fuel element (CFE) decreases and it gives rise to compact critical and equilibrium cores. The criticality studies of PARR-1 are in good agreement with the predictions

  10. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

    1997-01-01

    For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

  11. Research reactor core conversion programmes, Department of Research and Isotopes, International Atomic Energy Agency

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1985-01-01

    In order to put the problem of core conversion into perspective, statistical information on research reactors on a global scale is presented (from IAEA Research reactor Data Base). This paper describes the research reactor core conversion program of the Department of Research and Isotopes. Technical committee Meetings were held on the subject of research reactor core conversion since 1978, and results of these meetings are published in TECDOC-233, TECDOC-324, TECDOC-304. Additional publications are being prepared, several missions of experts have visited countries to discuss and help to plan core conversion programs; training courses and seminars were organised; IAEA has supported attendance of participants from developing countries to RERTR Meetings

  12. Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

    International Nuclear Information System (INIS)

    Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

    2009-01-01

    The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

  13. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    International Nuclear Information System (INIS)

    Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

    2004-01-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  14. Tools and applications for core design and shielding in fast reactors

    International Nuclear Information System (INIS)

    Rachamin, Reuven

    2013-01-01

    Outline: • Modeling of SFR cores using the Serpent-DYN3D code sequence; • Core shielding assessment for the design of FASTEF-MYRRHA; • Neutron shielding studies on an advanced Molten Salt Fast Reactor (MSFR) design

  15. Detectors for hydrodynamical processes in the reactor core

    International Nuclear Information System (INIS)

    Strube, D.

    1976-01-01

    The method described in this report is based on noise analytical measurements of electrical conductivity fluctuations of the cooling water stream in the reactor core. The conductivity fluctuations have their origin in steam and air bubbles and in cooling water mixing effects in regard to temperature and ionisation by gamma and neutron fields. The fluctuations are transformed into voltage signals by two electrodes in direction of the cooling water stream and then crosscorrelated. From the known distance of the two electrodes and the shift of the crosscorrelation function one can compute the velocity of the cooling medium and the bubbles. Void fractions were also determined with this detection device in out of pile experiments. (author)

  16. BR2 reactor core steady state transient modeling

    International Nuclear Information System (INIS)

    Makarenko, A.; Petrova, T.

    2000-01-01

    A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)

  17. Machine Learning of the Reactor Core Loading Pattern Critical Parameters

    Directory of Open Access Journals (Sweden)

    Krešimir Trontl

    2008-01-01

    Full Text Available The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm, and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper, we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employ a recently introduced machine learning technique, support vector regression (SVR, which is a data driven, kernel based, nonlinear modeling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modeling. We illustrate the performance of the solution and discuss its applicability, that is, complexity, speed, and accuracy.

  18. Oxidation during reflood of reactor core with melting cladding

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.; Allison, C.M.; Davis, K.L. [and others

    1995-09-01

    Models were recently developed and incorporated into the SCDAP/RELAP5 code for calculating the oxidation of fuel rods during cladding meltdown and reflood. Experiments have shown that a period of intense oxidation may occur when a hot partially oxidized reactor core is reflooded. This paper offers an explanation of the cladding meltdown and oxidation processes that cause this intense period of oxidation. Models for the cladding meltdown and oxidation processes are developed. The models are assessed by simulating a severe fuel damage experiment that involved reflood. The models for cladding meltdown and oxidation were found to improve calculation of the temperature and oxidation of fuel rods during the period in which hot fuel rods are reflooded.

  19. Reactor core with rod-shaped fuel cells

    International Nuclear Information System (INIS)

    Dworak, A.

    1975-01-01

    Power distribution in a high-temperature gas-cooled reactor is optimized. Especially the axial as well as the radial power distribution is kept constant, the core consisting of several consecutive rod-shaped fuel cells. To this end, the dwell times of the fuel cells are fitted to the given power distribution. Fuel cells with equal dwell times, seen in flow direction, are arranged side by side, and those with the shortest dwell times are placed in areas with the greatest power release. These areas ly on the coolant inlet side. To keep the power distribution constant, fuel cells with neutron poison or absorber rods with absorbing rates decreasing in flow direction can also be inserted. (RW/PB) [de

  20. Machine Learning of the Reactor Core Loading Pattern Critical Parameters

    International Nuclear Information System (INIS)

    Trontl, K.; Pevec, D.; Smuc, T.

    2008-01-01

    The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm, and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper, we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employ a recently introduced machine learning technique, support vector regression (SVR), which is a data driven, kernel based, nonlinear modeling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modeling. We illustrate the performance of the solution and discuss its applicability, that is, complexity, speed, and accuracy

  1. Diagnostic Technology Development for Core Internal Structure in CANDU reactor

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Cheong, Y. M.; Lee, Y. S. and others

    2005-04-01

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including measurement and monitoring technology has increased continuously. Because the fuel channels and the neighboring sensing tubes and control rods are particularly one of the critical components in CANDU nuclear plant, they are treated as a major research target in order to counteract the possible problems and establish the counterplan for the CANDU reactor safety improvement. To ensure the core structure integrity in CANDU nuclear plant, the following 2 research tasks were performed: Development of NDE technologies for the gap measurement between the fuel channels and LIN tubes. Development of vibration monitoring technology of the fuel channels and sensing tubes. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

  2. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    Nicolas, Anne

    1989-01-01

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated [fr

  3. Intelligent system for conceptural design of new reactor cores

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki

    1995-01-01

    The software system IRDS has been developed at Japan Atomic Energy Research Institute to support the conceptual design of a new type of reactor core in the fields of neutronics, thermohydraulics, and fuel behavior. IRDS involves various analysis codes, database, and man-machine interfaces that efficiently support a whole design process on a computer. The main purpose of conceptual design is to decide an optimal set of basic design parameters. Designers usually carry out many parametric survey calculations and search a design window (DW), which is a feasible parameter range satisfying design criteria and goals. An automatic DW search function is installed to support such works. The man-machine interface based on menu windows will enable nonspecialists to use various analysis codes easily

  4. Core reactivity estimation in space reactors using recurrent dynamic networks

    Science.gov (United States)

    Parlos, Alexander G.; Tsai, Wei K.

    1991-01-01

    A recurrent multilayer perceptron network topology is used in the identification of nonlinear dynamic systems from only the input/output measurements. The identification is performed in the discrete time domain, with the learning algorithm being a modified form of the back propagation (BP) rule. The recurrent dynamic network (RDN) developed is applied for the total core reactivity prediction of a spacecraft reactor from only neutronic power level measurements. Results indicate that the RDN can reproduce the nonlinear response of the reactor while keeping the number of nodes roughly equal to the relative order of the system. As accuracy requirements are increased, the number of required nodes also increases, however, the order of the RDN necessary to obtain such results is still in the same order of magnitude as the order of the mathematical model of the system. It is believed that use of the recurrent MLP structure with a variety of different learning algorithms may prove useful in utilizing artificial neural networks for recognition, classification, and prediction of dynamic systems.

  5. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    International Nuclear Information System (INIS)

    Hultqvist, Pontus

    2004-09-01

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

  6. Unsteady thermal analysis of gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    Lakkis, I.A.

    1993-01-01

    This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously

  7. Analytic function expansion nodal method for nuclear reactor core design

    International Nuclear Information System (INIS)

    Noh, Hae Man

    1995-02-01

    In most advanced nodal methods the transverse integration is commonly used to reduce the multi-dimensional diffusion equation into equivalent one- dimensional diffusion equations when derving the nodal coupling equations. But the use of the transverse integration results in some limitations. The first limitation is that the transverse leakage term which appears in the transverse integration procedure must be appropriately approximated. The second limitation is that the one-dimensional flux shapes in each spatial direction resulted from the nodal calculation are not accurate enough to be directly used in reconstructing the pinwise flux distributions. Finally the transverse leakage defined for a non-rectangular node such as a hexagonal node or a triangular node is too complicated to be easily handled and may contain non-physical singular terms of step-function and delta-function types. In this thesis, the Analytic Function Expansion Nodal (AFEN) method and its two variations : the Polynomial Expansion Nodal (PEN) method and the hybrid of the AFEN and PEN methods, have been developed to overcome the limitations of the transverse integration procedure. All of the methods solve the multidimensional diffusion equation without the transverse integration. The AFEN method which we believe is the major contribution of this study to the reactor core analysis expands the homogeneous flux distributions within a node in non-separable analytic basis functions satisfying the neutron diffusion equations at any point of the node and expresses the coefficients of the flux expansion in terms of the nodal unknowns which comprise a node-average flux, node-interface fluxes, and corner-point fluxes. Then, the nodal coupling equations composed of the neutron balance equations, the interface current continuity equations, and the corner-point leakage balance equations are solved iteratively to determine all the nodal unknowns. Since the AFEN method does not use the transverse integration in

  8. Investigation of activity release during light water reactor core meltdown

    International Nuclear Information System (INIS)

    Albrecht, H.; Matschoss, V.; Wild, H.

    1978-01-01

    A test facility was developed for the determination of activity release and of aerosol characteristics under realistic light water reactor core melting conditions. It is composed of a high-frequency induction furnace, a ThO 2 crucible system, and a collection apparatus consisting of membrane and particulate filters. Thirty-gram samples of a representative core material mixture (corium) were melted under air, argon, or steam at 0.8 to 2.2 bar. In air at 2700 0 C, for example, the relative release was 0.4 to 0.7% for iron, chromium, and cobalt and 4 to 11% for tin, antimony, and manganese. Higher release values of 20 to 40% at lower temperatures (2150 0 C, air) were found for selenium, cadmium, tellurium, and cesium. The size distribution of the aerosol particles was trimodal with maxima at diameters of 0.17, 0.30, and 0.73 μm. The result of a qualitative x-ray microanalysis was that the main elements of the melt were contained in each aerosol particle. Further investigations will include larger melt masses and the additional influence of concrete on the release and aerosol behavior

  9. A reactor core with accurately positioned fuel-batteries

    International Nuclear Information System (INIS)

    Borrman, B.E.

    1976-01-01

    A reactor core of containing a grid for a plurality of fuel batteries each of which is constituted by several claddings containing fuel-rods, said grid comprising square members mainly and being located at the core upper-end, each square member surrounding a group of four fuel batteries, spring-contacts being mounted between the fuel batteries and the grid, slots being provided between the batteries for the four arms of a centrally mounted cross-slaped control-rod, each slot being provided at the grid-level, with a flexible spacing device, the overall spacing of whork determining the (a+2b)- dimension is equal to, or higher than, the largest thickness of arm D of the above-mentioned control-rod, said spacing device constituting one of the control-rails the fuel batteries fixed to the fuel-element envelope, as well as the control-rails fixed to the grid, characterized in that each battery control-rail forms a closing surface at right angles to the wall of the adjacent battery and directed toward the grid nearest surface in contact with the above-mentioned control-rail. (author)

  10. Fast reactor core monitoring by analysis of temperature noise

    International Nuclear Information System (INIS)

    Dubuisson, B.; Smolarz, A.

    1984-01-01

    The study shows, with the results obtained, how it is possible to approach the problem of diagnosis with a technique based on the use of algorithms for statistical pattern recognition was justifiable. The results presented here, with a view to their use for fast breeder reactor core surveillance, are very encouraging, the most important point being the data representation. For this study, it was difficult to find the most suitable parameters for characterizing the various simulated core states, however, despite this handicap, the classification algorithm provided quite acceptable results. The second point concerns the characterization of a system's evolution. The criterion defined was chosen for adaptation to our algorithm. One acertained that it was possible to characterize evolution on the basis of this criterion as long as the rejected points were not too far from the known learning sets. Under these circumstances, the advantage in characterizing evolution in that the changes in evolution occur when the rejected points have a tendency to agglomerate in a small area of space could be seen. This phenomenon thus makes it possible to forsee whether the creation of a new class is possible. Where the rejected points are far away from the known learning sets, the criterion used proved to be too sensitive and the characterization of evolution was less satisfactory

  11. The neutron beam intensity increase by in-core fuel management enhancement in multipurpose research reactors

    International Nuclear Information System (INIS)

    Martinc, R.; Vukadin, Z.; Konstantinovic, J.

    1986-01-01

    The exploitation characteristics of an existing multipurpose research reactor can be increased not only by great reconstruction, but also, to the considerable extent, by the in-core fuel management sophistication. The optimisation of the in-core fuel management procedure in such reactors is governed (among others) by the identified reactor utilisation goals, i.e. by weighting factors dedicated to different utilisation goals, which are often (regarding the in-core fuel management procedure) highly controversial. In this work the best solution for in-core fuel management is sought, with the highest weighting factor dedicated to the neutron beam usage, rather than sample irradiation in the reactor core. The term in-core fuel management includes: the core configuration, the locations of the fresh fuel inflow zone and spent fuel excite zone, and the fuel transfers between these two zones (author)

  12. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  13. Reactor core with rod-shaped fuel cells

    International Nuclear Information System (INIS)

    Dworak, A.

    1977-01-01

    The aim is an optimization of load distribution in the core so that the load decreases in the direction of coolant flow (with gas cooling from above downwards) but so that it remains constant in horizontal layers to the edge of the core. The former produces optimum cooling, because the coolant has to take up decreasing heat output in the direction of flow. The latter simplifies refueling, because replacement of a whole layer having the same burn-up takes place. The upper two layers with the highest output and the shortest dwell time are replaced every 300 days, for example, the third layer is replaced after double this time and 5 more layers after four times this dwell time. After the simultaneous replacement of all layers, the reactor is in the same state as at commissioning. The fuel cells consist of hexagonal graphite blocks about 1.65 metres in height and 0.75 wide, for example. Each block contains about 100 through cooling channels and about 200 fuel channels closed on both sides. A large number of columns each consisting of 8 blocks is arranged in a tight honeycomb pattern and forms the core. Within each of the 8 horizontal layers of blocks, each fuel cell contains the same fuel mixture with predetermined dwell time. The fuel mixture is suited to the dwell time planned for each layer. The various fuel cells are kept at the same output by burnable neutron poisons in special channels provided for this purpose in the fuel cell and/or by absorber rods, or a planned load distribution is maintained. (HP) [de

  14. A nodal Grean's function method of reactor core fuel management code, NGCFM2D

    International Nuclear Information System (INIS)

    Li Dongsheng; Yao Dong.

    1987-01-01

    This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

  15. The influence of reactor core parameters on effective breeding coefficient Keff

    Institute of Scientific and Technical Information of China (English)

    Liu Li-Po; Liu Yi-Bao; Wang Juan; Yang Bo; Zhang Tao

    2008-01-01

    The values of effective breeding coefficient Keff in a reactor core of nuclear power plant are calculated for different values of parameters (core structure, fuel assembly component) by using the Monte Carlo method. The obtained values of Keff are compared and analysed, which can provide theoretical basis for reactor design.

  16. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated

  17. Innovative research reactor core designed. Estimation and analysis of gamma heating distribution

    International Nuclear Information System (INIS)

    Setiyanto

    2014-01-01

    The Gamma heating value is an important factor needed for safety analysis of each experiments that will be realized on research reactor core. Gamma heat is internal heat source occurs in each irradiation facilities or any material irradiated in reactor core. This value should be determined correctly because of the safety related problems. The gamma heating value is in general depend on. reactor core characteristics, different one and other, and then each new reactor design should be completed by gamma heating data. The Innovative Research Reactor is one of the new reactor design that should be completed with any safety data, including the gamma heating value. For this reasons, calculation and analysis of gamma heating in the hole of reactor core and irradiation facilities in reflector had been done by using of modified and validated Gamset computer code. The result shown that gamma heating value of 11.75 W/g is the highest value at the center of reactor core, higher than gamma heating value of RSG-GAS. However, placement of all irradiation facilities in reflector show that safety characteristics for irradiation facilities of innovative research reactor more better than RSG-GAS reactor. Regarding the results obtained, and based on placement of irradiation facilities in reflector, can be concluded that innovative research reactor more safe for any irradiation used. (author)

  18. Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    EL-Khawlani, Afrah [Physics Department, Sana' a (Yemen); Aziz, Moustafa [Nuclear and radiological regulatory authority, Cairo (Egypt); Ismail, Mahmud Yehia; Ellithi, Ali Yehia [Cairo Univ. (Egypt). Faculty of Science

    2015-03-15

    The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.

  19. Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

    International Nuclear Information System (INIS)

    Guelol Oezdere, Oya

    2003-01-01

    Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

  20. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  1. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  2. Estimation of a Reactor Core Power Peaking Factor Using Support Vector Regression and Uncertainty Analysis

    International Nuclear Information System (INIS)

    Bae, In Ho; Naa, Man Gyun; Lee, Yoon Joon; Park, Goon Cherl

    2009-01-01

    The monitoring of detailed 3-dimensional (3D) reactor core power distribution is a prerequisite in the operation of nuclear power reactors to ensure that various safety limits imposed on the LPD and DNBR, are not violated during nuclear power reactor operation. The LPD and DNBR should be calculated in order to perform the two major functions of the core protection calculator system (CPCS) and the core operation limit supervisory system (COLSS). The LPD at the hottest part of a hot fuel rod, which is related to the power peaking factor (PPF, F q ), is more important than the LPD at any other position in a reactor core. The LPD needs to be estimated accurately to prevent nuclear fuel rods from melting. In this study, support vector regression (SVR) and uncertainty analysis have been applied to estimation of reactor core power peaking factor

  3. Transient thermal characteristics of a core channel in a molten salt reactor

    International Nuclear Information System (INIS)

    Sakashita, H.; Ishiguro, R.; Sugiyama, K.

    1987-01-01

    The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

  4. An analysis of cobalt irradiation in CANDU 6 reactor core

    International Nuclear Information System (INIS)

    Gugiu, E.D.; Dumitrache, I.

    2003-01-01

    In CANDU reactors, one has the ability to replace the stainless steel adjuster rods with neutronically equivalent Co assemblies with a minimum impact on the power plant safety and efficiency. The 60 Co produced by 59 Co irradiation is used extensively in medicine and industry. The paper mainly describes some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronically equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and the heating of the irradiated cobalt rods are performed using the Monte Carlo codes MCNP5 and MONTEBURNS2.1. The 60 Co activity and heating evaluations are closely related to the neutronics computations and to the density evolution of cobalt isotopes during assumed in-core irradiation period. Unfortunately, the activities of these isotopes could not be evaluated directly using the burn-up capabilities of the MONTEBURNS code because of the lack of their neutron cross-section from the MCNP5 code library. Additional MCNP5 runs for all the cobalt assemblies have been done in order to compute the flux-spectrum, the 59 Co and the 60 Co radiative capture reaction rates in the adjusters. The 60m Co cross-section was estimated using the flux-spectrum and the ORIGEN2.1 code capabilities THERM and RES. These computational steps allowed the evaluation of the one-group cross-section for the radiative capture reactions of cobalt isotopes. The values obtained replaced the corresponding ones from the ORIGEN library, which have been estimated using the flux-spectrum specific to the fuel. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. (authors)

  5. Fuel density effect on parameter of reactivity coefficient of the Innovative Research Reactor core

    International Nuclear Information System (INIS)

    Rokhmadi; Tukiran S

    2013-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR) and then as an exchanger for old research reactor. The aim of this research is to complete RRI core design data as a requirement for design license. Calculation done is to get the RRI core reactivity coefficients with 5 x 5 core configuration and 20 MW of power, has more than 40 days cycle of length. The RRI core reactivity coefficient calculation is done for new U-"9Mo-Al fuel with variation of densities. The calculation is done by using WIMSD-5B and BATAN-FUEL computer codes. The result of calculation for conceptual design showed that the equilibrium RRI core with 5 x 5 configuration, 450 g, 550 g and 700 g of fuel loadings have negative reactivity coefficients of fuel temperature, moderator temperature, void fraction and density of moderator but the values of the reactivities are very variation. This results has met the safety criteria for RRI core conceptual design. (author)

  6. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  7. Contributions to the determination of the thermal core reliability of pressurized water reactors

    International Nuclear Information System (INIS)

    Ackermann, G.; Horche, W.; Melchior, H.; Prasser, H.M.

    1982-09-01

    The investigations in the field of thermohydraulics of PW reactors are aimed at a possible increase of economy and reliability of WWER-type-reactors. In detail the flow distribution at the core entrance, the modification of the power distribution as a result of an irregular temperature distribution at the core entrance, and based on the theory of hot spots the thermic core reliability are studied. In this connection qualitatively new methods are applied characterized by low expenditure. (author)

  8. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  9. Analysis of stress in reactor core vessel under effect of pressure lose shock wave

    International Nuclear Information System (INIS)

    Li Yong; Liu Baoting

    2001-01-01

    High Temperature gas cooled Reactor (HTR-10) is a modular High Temperature gas cooled Reactor of the new generation. In order to analyze the safety characteristics of its core vessel in case of large rupture accident, the transient performance of its core vessel under the effect of pressure lose shock wave is studied, and the transient pressure difference between the two sides of the core vessel and the transient stresses in the core vessel is presented in this paper, these results can be used in the safety analysis and safety design of the core vessel of HTR-10. (author)

  10. Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

    International Nuclear Information System (INIS)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin

    2016-01-01

    CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future

  11. Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

  12. Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future.

  13. Device for judging the stability of a reactor core

    International Nuclear Information System (INIS)

    Tanisaka, Satoshi; Fukunishi, Koyu.

    1980-01-01

    Purpose: To enable early detection for the abnormal states affecting on the core stability, by the extraction of a delay time in a self-correlation function as the characteristic value for the self-correlation function of reactor power. Constitution: A self-correlation function is always calculated in a self-correlation function operator and the calculated results are converted in a D/A converter into analog values. Negative voltage level in the values is extracted to an amplifier, then inputted to a peak-hold circuit and inputted into a differential amplifier and the output signal is inputted to a discriminator. The discriminator converts the input signal into binary voltage level signals, for example, into 0 volt and +5 volt at the output. The operation start of the differentiator is triggered by the binary signal. The input signals to the differentiator include a signal corresponding to the delay time when the negative voltage level stored as the normal state in the memory unit takes the maximum value and a signal corresponding to the delay time when the voltage level for the continuously calculated self-correlation function takes the maximum value. When the differential value between them exceeds a set value, an alarming device issues alarms which are stored in the recorder. (Seki, T.)

  14. Modelling Pressurized Water Reactor cores in terms of porous media

    International Nuclear Information System (INIS)

    Ricciardi, G.; Collard, B.; Ricciardi, G.; Bellizzi, S.; Cochelin, B.

    2009-01-01

    The aim of this study is to develop a tractable model of a nuclear reactor core taking the complexity of the structure (including its nonlinear behaviour) and fluid flow coupling into account. The mechanical behaviour modelling includes the dynamics of both the fuel assemblies and the fluid. Each rod bundle is modelled in the form of a deformable porous medium; then, the velocity field of the fluid and the displacement field of the structure are defined over the whole domain. The fluid and the structure are first modelled separately, before being linked together. The equations of motion for the structure are obtained using a Lagrangian approach and, to be able to link up the fluid and the structure, the equations of motion for the fluid are obtained using an arbitrary Lagrangian Eulerian approach. The finite element method is applied to spatially discretize the equations. Simulations are performed to analyse the effects of the characteristics of the fluid and of the structure. Finally, the model is validated with a test involving two fuel assemblies, showing good agreement with the experimental data. (authors)

  15. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

  16. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... comments were received. A companion guide, DG-1277, ``Initial Test Program of Emergency Core Cooling... NUCLEAR REGULATORY COMMISSION [NRC-2011-0129] Preoperational Testing of Emergency Core Cooling... (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

  17. The system of the measurement of reactor power and the monitoring of core power distribution

    International Nuclear Information System (INIS)

    Li Xianfeng

    1999-01-01

    The author mainly describes the measurement of the reactor power and the monitoring of the core power distribution in DAYA BAY nuclear power plant, introduces the calibration for the measurement system. Ex-core nuclear instrumentation system (RPN) and LOCA surveillance system (LSS) are the most important system for the object. they perform the measurement of the reactor power and the monitoring of the core power distribution on-line and timely. They also play the important roles in the reactor control and the reactor protection. For the same purpose there are test instrumentation system (KME) and in-core instrumentation system (RIC). All of them work together ensuring the exact measurement and effective monitoring, ensuring the safety of the reactor power plant

  18. Assessment of core protection and monitoring systems for an advanced reactor SMART

    International Nuclear Information System (INIS)

    In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

    2002-01-01

    Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

  19. Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

    International Nuclear Information System (INIS)

    Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

    2012-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

  20. Investigation of Equilibrium Core by recycling MA and LLFP in fast reactor cycle (I)

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Shono, Akira; Ishikawa, Makoto

    1999-05-01

    Feasibility study on a self-consistent fuel cycle system is performed in the nuclear fuel recycle system with fast reactors. In this system, the self-generated MAs (Minor Actinides) and LLFPs (Long Lived Fission Products) are confined and incinerated in the fast reactor. Analyses of the nuclear properties for an 'Equilibrium Core', in which the self-generated MAs and LLFPs are confined, are investigated. A conventional sodium cooled oxide fuel fast reactor is selected as the core specifications for the 'Equilibrium Core'. This 600 MWe fast reactor does not have a radial blanket. In this study, the nuclear characteristics of the 'Equilibrium Core' are compared with those of a 'Standard Core' and '5 w/oMA Core'. The 'Standard Core' does not confine MAs and LLFPs in the core, and a 5 w/o-MA Rom LWR is loaded in the '5 w/oMA Core'. Through this comparison between 'Equilibrium Core' and the others, the specific characters of the 'Equilibrium Core' are investigated. In order to realize the 'Equilibrium Core' in the viewpoint of nuclear properties, whether the conventional design concept of fast reactors must be changed or not is also evaluated. The analyses for the nitride and metallic fuel cores are also performed because of their different nuclear characteristics compared with the oxide fuel core. Assuming the separation of REs (Rare Earth elements) from MAs and the isotope separation of LLFPs, most of the nuclear properties for the 'Equilibrium Core' are not beyond those for the '5 w/oMA Core'. It is, therefore, possible to bring the 'Equilibrium Core' into existence without any drastic modification for the design concept of the typical oxide fuel fast reactors. Although the 15.1[w/o] LLFPs are loading in the core of the oxide fuel 'Equilibrium Core', a breeding ratio is more than 1.0 and the difference in a amount of plutonium between a charging and discharging is only 0.04 [ton/year]. Without any drastic change for the design concept of the conventional oxide fuel

  1. Chemical and X-ray diffraction analysis on selected samples from the TMI-2 reactor core

    International Nuclear Information System (INIS)

    Kleykamp, H.; Pejsa, R.

    1991-05-01

    Selected samples from different positions of the damaged TMI-2 reactor core were investigated by X-ray microanalysis and X-ray diffraction. The measurements yield the following resolidified phases after cooling: Cd and In depleted Ag absorber material, intermetallic Zr-steel compounds, fully oxidized Zircaloy, UO 2 -ZrO 2 solid solutions and their decomposed phases, and Fe-Al-Cr-Zr spinels. The composition of the phases and their lattice parameters as well as the eutectic and monotectic character can serve as indicators of local temperatures of the core. The reaction sequences are estimated from the heterogeneous equilibria of these phases. The main conclusions are: (1) Liquefaction onset is locally possible by Inconel-Zircaloy and steel-Zircaloy reactions of spacers and absorber guide tubes at 930deg C. However, increased rates of dissolution occur above 1200deg C. (2) UO 2 dissolution in the Inconel-steel-Zircaloy melt starts at 1300deg C with increased rates above 1900deg C. (3) Fuel temperatures in the core centre are increased above 2550deg C, liquid (U,Zr)O 2 is generated. (4) Square UO 2 particles are reprecipitated from the Incoloy-steel-Zircaloy-UO 2 melt during cooling, the remaining metallic melt is oxygen poor; two types of intermetallic phases are formed. (5) Oxidized Fe and Zr and Al 2 O 3 from burnable absorber react to spinels which form a low melting eutectic with the fuel at 1500deg C. The spinel acts as lubricant for fuel transport to the lower reactor plenum above 1500deg C. (6) Ruthenium (Ru-106) is dissolved in the steel phase, antimony (Sb-125) in the α-Ag absorber during liquefaction. (7) Oxidation of the Zircaloy-steel phases takes place mainly in the reflood stage 3 of the accident scenario. (orig.) [de

  2. Development of conceptual nuclear design of 10MWt research reactor core

    International Nuclear Information System (INIS)

    Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

    2008-03-01

    KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

  3. Selection method and device for reactor core performance calculation input indication

    International Nuclear Information System (INIS)

    Yuto, Yoshihiro.

    1994-01-01

    The position of a reactor core component on a reactor core map, which is previously designated and optionally changeable, is displayed by different colors on a CRT screen by using data of a data file incorporating results of a calculation for reactor core performance, such as incore thermal limit values. That is, an operator specifies the kind of the incore component to be sampled on a menu screen, to display the position of the incore component which satisfies a predetermined condition on the CRT screen by different colors in the form of a reactor core map. The position for the reactor core component displayed on the CRT screen by different colors is selected and designated on the screen by a touch panel, a mouse or a light pen, thereby automatically outputting detailed data of evaluation for the reactor core performance of the reactor core component at the indicated position. Retrieval of coordinates of fuel assemblies to be data sampled and input of the coordinates and demand for data sampling can be conducted at once by one menu screen. (N.H.)

  4. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Rimpault, G.

    2004-01-01

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO 2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  5. Design of a reactor core in the Oma Full MOX-ABWR

    International Nuclear Information System (INIS)

    Hama, Teruo

    1999-01-01

    The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

  6. Improvements to the sodium supply system of a nuclear reactor core

    International Nuclear Information System (INIS)

    Chevallier, Rene; Marchais, Christian.

    1981-01-01

    This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

  7. Utilization of local area network technology and decentralized structure for nuclear reactor core temperature monitoring

    International Nuclear Information System (INIS)

    Casella, M.; Peirano, F.

    1986-01-01

    The present system concerns Superphenix type reactors. It is a new version of system for monitoring the reactor core temperatures. It has been designed to minimize the cost and the wiring complexity because of the large number of channels (800). For this, equipments are arranged on the roof slab of the reactor with a single link to the control room; from which the name Integrated Treatment of Core Temperatures: TITC 1500 and the natural choice of a distributed system. This system monitors permanently the thermal state of the core a Superphenix type reactor. This monitoring system aims at detecting anomalies of core temperature rise, releasing automatic shutdown (safety), and providing to the monitoring systems not concerned safety the information concerning the core [fr

  8. Aspects of unconventional cores for large sodium cooled power reactors; evaluation of a literature survey

    International Nuclear Information System (INIS)

    Kiefhaber, E.

    1978-10-01

    The report gives an overview of a literature study on the application of unconventional cores for sodium cooled fast reactors. Different types of unconventional cores (heterogeneous cores, pancake cores, moderated cores and others) are compared with conventional cores, which are characterized by a cylindrical geometry with two or three fissile zones surrounded by an axial and a radial blanket. The main parameters of interest in this comparison are the neutronic parameters sodium void and Doppler effect, the breeding properties and the steel damage. Consequences for the core safety and the overall plant design are also mentioned

  9. Fault current limiter-predominantly resistive behavior of a BSCCO shielded-core reactor

    International Nuclear Information System (INIS)

    Ennis, M. G.; Tobin, T. J.; Cha, Y. S.; Hull, J. R.

    2000-01-01

    Tests were conducted to determine the electrical and magnetic characteristics of a superconductor shielded core reactor (SSCR). The results show that a closed-core SSCR is predominantly a resistive device and an open-core SSCR is a hybrid resistive/inductive device. The open-core SSCR appears to dissipate less than the closed-core SSCR. However, the impedance of the open-core SSCR is less than that of the closed-core SSCR. Magnetic and thermal diffusion are believed to be the mechanism that facilitates the penetration of the superconductor tube under fault conditions

  10. Method and apparatus for monitoring the axial power distribution within the core of a nuclear reactor, exterior of the reactor

    International Nuclear Information System (INIS)

    Graham, K.F.; Gopal, R.

    1978-01-01

    A method and apparatus for establishing the axial flux distribution of a reactor core from monitored responses obtained exterior of the reactor is described. The monitored responses are obtained from at least three axially spaced flux responsive detectors that are positioned within proximity of the periphery of the reactor core. The detectors provide corresponding electrical outputs representative of the flux monitored. The axial height of the core is figuratively divided at a plurality of space coordinates sufficient to provide reconstruction in point representation of the relative flux shape along the core axis. The relative value of flux at each of the spaced coordinates is then established from a sum of the electrical outputs of the detectors, respectively, algebraically modified by a corresponding preestablished constant

  11. Split core experiments; Part I. Axial neutron flux distribution measurements in the reactor core with a central horizontal reflector

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Raisic, N; Obradovic, D; Jovanovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-05-01

    A series of critical experiments were performed on the RB reactor in order to determine the thermal neutron flux increase in the central horizontal reflector formed by a split reactor core. The objectives of these experiments were to study the possibilities of improving the thermal neutron flux characteristics of the neutron beam in the horizontal beam tube of the RA research reactor. The construction of RA reactor enables to split the core in two, to form a central horizontal reflector in front of the beam tube. This is achieved by replacing 2% enriched uranium slugs in the fuel channel by dummy aluminium slugs. The purpose of the first series of experiments was to study the gain in thermal neutron component inside the horizontal reflector and the loss of reactivity as a function of the lattice pitch and central reflector thickness.

  12. Feasibility study to restart the research reactor RA with a converted core

    International Nuclear Information System (INIS)

    Matausek, M.V.; Plecas, I.; Marinkovic, N.

    1999-01-01

    Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities, are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

  13. Design study of eventual core conversion for the research reactor RA

    International Nuclear Information System (INIS)

    Matausek, M. V.; Marinkovic, N.

    1998-01-01

    Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective, multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

  14. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  15. Monte Carlo applications to core-following of the National Research Universal reactor (NRU)

    International Nuclear Information System (INIS)

    Nguyen, T.S.; Wang, X.; Leung, T.

    2014-01-01

    Reactor code TRIAD, relying on a two-group neutron diffusion model, is currently used for core-following of NRU - to track reactor assembly locations and burnups. The Monte Carlo (MCNP or SERPENT) full-reactor models of NRU can be used to provide the core power distribution for calculating fuel burnups, with WIMS-AECL providing fuel depletion calculations. The MCNP/WIMS core-following results were in good agreement with the measured data, within the expected biases. The Monte Carlo methods, still very time-consuming, need to be able to run faster before they can replace TRIAD for timely support of NRU operations. (author)

  16. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

    Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

    2013-01-01

    Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

  17. Irradiation-accelerated corrosion of reactor core materials

    International Nuclear Information System (INIS)

    Bartels, David; Was, Gary; Jiao, Zhijie

    2012-09-01

    The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, but also applies to most all other GenIV concepts. Of these four drivers, the combination of radiation and corrosion presents a unique and extremely challenging environment for materials, for which an understanding of the fundamental science is essentially absent. Irradiation can affect corrosion or oxidation in at least three different ways. Radiation interaction with water results in the decomposition of water into radicals and oxidizing species that will increase the electrochemical corrosion potential and lead to greater corrosion rates. Irradiation of the solid surface can produce excited states that can alter corrosion, such as in the case of photo-induced corrosion. Lastly, displacement damage in the solid will result in a high flux of defects to the solid-solution interface that can alter and perhaps, accelerate interface reactions. While there exists reasonable understanding of how corrosion is affected by irradiation of the aqueous environment, there is little understanding of how irradiation affects corrosion through its impact on the solid, whether metal or oxide. The reason is largely due to the difficulty of conducting experiments that can measure this effect separately. We have undertaken a project specifically to separate the several effects of irradiation on the mechanisms of corrosion. We seek to answer the question: How does radiation damage to the solution-oxide couple affect the oxidation process differently from radiation damage to either component alone? The approach taken in this work is to closely compare corrosion accelerated by (1) proton irradiation, (2) electron irradiation, and (3) chemical corrosion potential effects alone, under typical PWR operating conditions at 300 deg. C. Both 316 stainless steel and zirconium are to be studied. The proton

  18. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations

  19. Adaptive control method for core power control in TRIGA Mark II reactor

    Science.gov (United States)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  20. In-core fuel management for the course on operational physics of power reactors

    International Nuclear Information System (INIS)

    Levine, S.H.

    1982-01-01

    The heart of a nuclear power station is the reactor core producing power from the fissioning of uranium or plutonium fuel. Expertise in many different technical fields is required to provide fuel for continuous economical operation of a nuclear power plant. In general, these various technical disciplines can be dichotomized into ''Out-of-core'' and ''In-core'' fuel management. In-core fuel management is concerned, as the name implies, with the reactor core itself. It entails calculating the core reactivity, power distribution, and isotopic inventory for the first and subsequent cores of a nuclear power plant to maintain adequate safety margins and operating lifetime for each core. In addition, the selection of reloading schemes is made to minimize energy costs

  1. Operational report, Formation of the XXVII reactor core, plan of fuel exchange

    International Nuclear Information System (INIS)

    Martinc, R.

    1977-01-01

    Plan for fuel exchange for formation of the reactor core No. XXVII is presented. This report includes: the quantity of 80% enriched fuel which is input in the core, description of the fuel 'transfer' through the core within this fuelling scheme. It covers the review of reactor safety operating with the core No. XXVII related to reactivity change, thermal load of the fuel channels and fuel burnup. These data result from the analysis based on the same correlated calculation method which was applied for planning the first regular fuel exchange with 80% enriched fuel (core No. XXVI configuration), which has been approved in february 1977. Based on the enclosed data and the fuel exchange according to the proposed procedure it is expected that the reactor operation with core No. XXVII configuration will be safe [sr

  2. Optimal power and distribution control for weakly-coupled-core reactor

    International Nuclear Information System (INIS)

    Oohori, Takahumi; Kaji, Ikuo

    1977-01-01

    A numerical procedure has been devised for obtaining the optimal power and distribution control for a weakly-coupled-core reactor. Several difficulties were encountered in solving this optimization problem: (1) nonlinearity of the reactor kinetics equations; (2) neutron-leakage interaction between the cores; (3) localized power changes occurring in addition to the total power changes; (4) constraints imposed on the states - e.g. reactivity, reactor period. To obviate these difficulties, use is made of the generalized Newton method to convert the problem into an iterative sequence of linear programming problems, after approximating the differential equations and the integral performance criterion by a set of discrete algebraic equations. In this procedure, a heuristic but effective method is used for deriving an initial approximation, which is then made to converge toward the optimal solution. Delayed-neutron one-group point reactor models embodying transient temperature feed-back to the reactivity are used in obtaining the kinetics equations for the weakly-coupled-core reactor. The criterion adopted for determining the optimality is a norm relevant to the deviations of neutron density from the desired trajectories or else to the time derivatives of the neutron density; uniform control intervals are prescribed. Examples are given of two coupled-core reactors with typical parameters to illustrate the results obtained with this procedure. A comparison is also made between the coupled-core reactor and the one-point reactor. (auth.)

  3. Conceptual core designs for a 1200 MWe sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

    2008-01-01

    The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

  4. Development of Structural Core Components for Breeder Reactors

    International Nuclear Information System (INIS)

    Saibaba, N.

    2013-01-01

    Core structural materials: • The desire is to have only fuel in the core, structural material form 25% of the total core: – To support and to retain the fuel in position; – Provide necessary ducts to make coolant flow through & transfer/remove heat. • For 500 MWe FBR with Oxide fuel (Peak Linear Power 450 W/cm), total fuel pins required in the core are of the order 39277 pins (both inner & outer core Fuel SA); • Considering 217 pins/Fuel SA there are 181 Fuel SA wrapper tubes • These structural materials see hostile core with max temperature and neutron flux

  5. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2011-01-01

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  6. Solid state laser driver for an ICF reactor

    International Nuclear Information System (INIS)

    Krupke, W.F.

    1988-01-01

    A conceptual design is presented of the main power amplifier of a multi-beamline, multi-megawatt solid state ICF reactor driver. Simultaneous achievement of useful beam quality and high average power is achieved by a proper choice of amplifier geometry. An amplifier beamline consists of a sequence of face-pumped rectangular slab gain elements, oriented at the Brewster angle relative to the beamline axis, and cooled on their large faces by helium gas that is flowing subsonically. The infrared amplifier output radiation is shifted to an appropriately short wavelength ( 10% (including all flow cooling input power) when the amplifiers are pumped by efficient high-power AlGaAs semiconductor laser diode arrays. 11 refs., 3 figs., 7 tabs

  7. Reactivity variations associated with the core expansion of the MARIA research reactor after modernisation

    International Nuclear Information System (INIS)

    Krzysztoszek, G.

    1997-01-01

    Polish high flux research reactor MARIA is a pool type reactor moderated with beryllium and water and cooled with water. The fuel is 80% enriched uranium, in the shape of multitube fuel elements, each tube made up of UAl x alloy in aluminium cladding. MARIA reactor has been operated in the years of 1977-85 and then it was modernised and again put into operation in December 1992. The modernisation as regarded the reactor core comprises a beryllium matrix expansion from 20-48 blocks. Within the frame of the power start-up and trial operation the reactor has been extended from 12 to 18 fuel channels. On that stage of reactor operation the power of mostly loaded fuel channels was constrained to 1,6 MW. Reactor has been operated within the 100-hrs campaign for an irradiation of target materials and for performing measurements at the horizontal channel outlets. In the previous time it has been noticed substantial differences in reactivity changes of the core in similar campaigns of reactor operation. It concerns the reactivity losses during poisoning period of the reactor within the first 30-40 hrs of operation as well as in the fuel burning up process. An analysis of the reactivity variations during the core extension will made possible the fuel management optimisation in further reactor operation system. (author)

  8. The in-core experimental program at the MIT Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kohse, G.E.; Hu, L-W., E-mail: kohse@mit.edu [Massachusetts Inst. of Technology, Nuclear Reactor Lab., Cambridge, Massachusetts (United States)

    2014-07-01

    This paper describes the program of in-core experiments at the Massachusetts Institute of Technology Research Reactor (MITR), a 6 MW research reactor. The MITR has a neutron flux and spectrum similar to those in water-cooled power reactors and therefore provides a useful test environment for materials and fuels research. In-core facilities include: a water loop operating at pressurized water or boiling water reactor conditions, an inert gas irradiation facility operating at temperature up to 850 {sup o}C and special purpose facilities including fuel irradiation experiments. Recent and ongoing tests include: water loop investigations of corrosion and thermal and mechanical property evolution of SiC/SiC composites for fuel cladding, irradiation of advanced materials and in-core sensors at elevated temperatures, irradiation in molten fluoride salt at 700 {sup o}C of metal alloy, graphite and composite materials for power reactor applications and instrumented irradiations of metal-bonded hydride fuel. (author)

  9. A safety design approach for sodium cooled fast reactor core toward commercialization in Japan

    International Nuclear Information System (INIS)

    Kubo, Shigenobu

    2012-01-01

    JAEA’s safety approach for SFR core design is based on defence‐in‐depth concept, which includes DBAs and DECs (prevention and mitigation): • The reactor core is designed to have inherent reactivity feedback characteristics with negative power coefficient. • Operation temperature range is set sufficiently below the coolant boiling temperature so as to avoid coolant boiling against anticipated operational occurrences and DBAs. • If the plant state deviates from operational states, the safe reactor shutdown is achieved by automatic insertion of control rods. 2 active reactor shutdown systems are provided. • Failure of active reactor shutdown is assumed in a design extension condition . Passive shutdown capability is provided by SASS under such condition. • As a design extension condition, core disruptive accident is assumed. In order to prevent severe mechanical energy release which might cause containment function failure, core sodium void worth is limited below 6 dollars and molten fuel discharge capability is utilized by FAIDUS. (author)

  10. An overview of the Indian program related to fast reactor core mechanical behaviour

    International Nuclear Information System (INIS)

    Govindarajan, S.; Bhoje, S.B.; Paranjpe, S.R.

    1984-01-01

    This Indian review paper presents the evolution of the fast breeder program which began with fast breeder test reactor (FBTR) commencing in 1972. The state-of-art in the field of core mechanical behaviour is reviewed

  11. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    Tsuiki, Makoto; Sakurada, Koichi; Yoshida, Hiroyuki.

    1988-01-01

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  12. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

  13. Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1985-01-01

    A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

  14. Nuclear reactor with a fixed system of neutron poison, which can be burnt up, introduced into the reactor core

    International Nuclear Information System (INIS)

    Mueller, E.; Roegler, H.J.; Wickert, M.

    1985-01-01

    The fixed system consists of neutron poison which can be burnt up, in an uneven distribution, and with adjustable absorber rods for output control, which are driven into the reactor core from the side along the fuel elements. There is an excess of neutron poison which can be burnt up, overall, on the side of the reactor core away from the absorber rods. The reactor core is free of neutron poison which can be burnt up on the side where the absorber rods are driven in, so that the ratio of maximum to mean power density with reference to a possible absorber rod positions is less than for homogeneous distribution of the neutron poison which can be burnt up. (orig./HP) [de

  15. An analysis of reactor structural response to fuel sodium interaction in a hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

    1976-01-01

    This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A., calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. This work was supported by a grant from Power Reactor and Nuclear Fuel Development Corporation. (auth.)

  16. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  17. Gas-solid hydroxyethylation of potato starch in a stirred vibrating fluidized bed reactor

    NARCIS (Netherlands)

    Kuipers, N.J M; Stamhuis, Eize; Beenackers, A.A C M

    A novel reactor for modifying cohesive C-powders such as in the gas-solid hydroxyethylation of semidry potato starch is characterized, the so-called stirred vibrating fluidized bed reactor. Good fluidization characteristics are obtained in this reactor for certain combinations of stirring and

  18. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  19. Various reactivity effects value for assuring fast reactor core inherent safety

    International Nuclear Information System (INIS)

    Belov, S.B.; Vasilyev, B.A.

    1991-01-01

    The paper presents the results of temperature and power reactivity feedback components calculations for fast reactors with different core volume when using oxide, carbide, nitride and metal fuel. Reactor parameters change in loss of flow without scram and transient over power without scram accidents was evaluated. The importance of various reactivity feedback components in restricting the consequences of these accidents has been analyzed. (author)

  20. Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    D'Utra Bitelli, U.

    1993-01-01

    This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

  1. A Conceptual Supercritical Water Cooled Reactor Design Using a Cruciform Solid Moderator

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Bae, Kang Mok; Yoo, Jae Woon; Lee, Hyun Chul; Noh, Jae Man; Bae, Yoon Yong

    2006-01-15

    A Super Critical Water-Cooled Reactor(SCWR) concept proposed by Gen-IV has an advantage of a high thermal efficiency. However, there are some difficulties in neutronic core design for a SCWR due to lower moderator density resulting from the high operating temperature over the pseudo-critical temperature. In this report, the design concepts for the fuel assembly and the core for a SCWR were described as a feasibility study on the SCWR core design. HELIOS lattice code which will be used for group constants generation was verified for the application to the low coolant density condition of a SCWR. The TAF module for a thermal hydraulic feedback in MASTER was modified to consider high pressure and temperature of the supercritical coolant with single-phase fluid. A cruciform ZrH{sub 2} solid moderator was proposed for the SCWR fuel assembly design to compensate the lower coolant density. The axial zoning concept with three different enrichments for a fuel rod was used for the axial power shape control. Gadolinia burnable poison rods were used to reduce excess reactivity. Control rod system was grouped into 6 banks to control the excess reactivity of the core during normal operation. An orifice concept for each assembly was applied to control a coolant flow rate individually. As a result of the neutronic analysis for the equilibrium SCWR core, the maximum linear heat generation rete limit was satisfied and the maximum coolant temperature of the core outlet was {approx}590 .deg. C which is lower than 620 .deg. C of the maximum clad temperature limit.

  2. Influence of high dose irradiation on core structural and fuel materials in advanced reactors

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

  3. Pebble Bed Reactor: core physics and fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Worley, B.A.

    1979-10-01

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

  4. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

  5. Analysis of fuel management pattern of research reactor core of the MTR type design

    International Nuclear Information System (INIS)

    Lily Suparlina; Tukiran Surbakti

    2014-01-01

    Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x10 15 n/cm 2 s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 10 15 n/cm 2 s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

  6. Sampling of reactor pressure vessel and core internals

    International Nuclear Information System (INIS)

    Oberhaeuser, Ralf

    2012-01-01

    Decommissioning and dismantling of nuclear power plants is a growing business as a huge number of plants built in the 1970's have now reached their lifetime. It is well known that dismantling a nuclear power plant means an extraordinary expense for the owner respectively operator. Beside the dismantling works for itself, the disposal of activated components and other nuclear waste is very expensive. What comes next is the fact that final disposal facilities are not available yet in most countries meaning a need for interim storage on-site in specially built facilities. It can be concluded that a special attention is paid on producing a minimal radioactive waste volume. For this, optimized dismantling and packaging concepts have to be developed. AREVA is proud of versatile experience in successfully dismantling nuclear components like core internals and reactor pressure vessel (RPV). The basis of a well-founded and optimized dismantling and packaging concept must always be the detailed knowledge of the radiological condition of the component to be and in the best case a 3D activation- model. For keeping the necessary sampling effort as small as possible, but simultaneously as efficient as possible, representative sampling positions are defined in advance by theoretical radiological examinations. For this, a detailed 3D-CAD-model of the components to be dismantled has proven very helpful and effective. Under these aspects a sampling of RPV and its components is necessary to verify the theoretically calculated radiological data. The obtained results of activation and contamination are taken into account for the optimized dismantling and packaging strategy. The precise 3D-activation-model will reduce the necessary number and type of final disposal containers as security factors are minimized leading to a lower shielding effort, too. Besides, components or even parts of components may be subject of release measurement. In the end, costs can be reduced. In this context

  7. Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

    International Nuclear Information System (INIS)

    Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

    2008-01-01

    In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

  8. Development and study of a control and reactor shutdown device for FBR-type reactors with a modified open core

    International Nuclear Information System (INIS)

    Goswami, S.

    1983-01-01

    The doctoral thesis at hand presents a newly designed control and shutdown device to be used for output control and fast shutdown of modified open core FBR-type reactors. The task was the design of a new control and shutdown device having economic and operation advantages, using reactor components time-tested under reactor conditions. This control and shutdown device was adapted to the specific needs concerning dimensions and design. The actuation is based on the magnetic-jack principle, which has been upgraded for the purpose. The principle is now combined with pneumatic acceleration. The improvements mainly concern a smaller number of piece parts and system simplification. (orig./RW) [de

  9. Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs. Final summary report

    International Nuclear Information System (INIS)

    Greenspan, E

    2006-01-01

    The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity in particular for BWR's, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR's and BWR's without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR's and BWR's were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density ? on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR's more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fueled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ∼2/3 that of the MOX fuel and the discharged hydride fuel is

  10. Development of Core Design Model for Small-Sized Research Reactor and Establishment of Infrastructure for Reactor Export

    International Nuclear Information System (INIS)

    Kim, M. H.; Win, Naing; Lim, J. Y.

    2007-02-01

    Within 10 years a growing world-wide demand of new research reactor construction is expected because of obsolescence. In Korea, a new research reactor is also required in order to meet domestic demand of utilization. KAERI has been devoted to develop an export-oriented research reactors for these kinds of demand. A next generation research reactor should comply with general requirements for safety, economics, environment-friendliness and non-proliferation as well as high performance requirement of high flux level. A export-tailored reactor should be developed for the demand of developing counties or under-developed countries. A new design concept is to be developed for a long cycle length core which has excellent irradiation facility with high flux

  11. Contribution to the use of a solid moderator gas reactor, for naval propulsion

    International Nuclear Information System (INIS)

    Pheline, J.; Gautier, A.

    1960-01-01

    In this contribution, the authors discuss works performed in France for the development of nuclear propulsion in merchant ships, notably for an oil tanker of 50.000 tons with 17 knot speed, i.e. a 20.000 Hp engine with an energy produced by a 60 MW gas reactor with a solid moderator and comprising 400 channels loaded with uranium oxide enriched ay 2.8 per cent and sheathed with a refractory alloy. The authors discuss the possible materials for the moderator, the heat transfer medium, the sheath, the fuel and the structures, and report technological studies (mechanical tests, irradiation tests) performed to investigate material properties and their behaviour in operation conditions. They report tests performed to investigate core structure characteristics with respect to neutrons. They finally briefly present a prototype

  12. SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

    Science.gov (United States)

    Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

    2016-04-01

    Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

  13. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

  14. Simulation an Accelerator driven Subcritical Reactor core with thorium fuel

    International Nuclear Information System (INIS)

    Shirmohammadi, L.; Pazirandeh, A.

    2011-01-01

    The main purpose of this work is simulation An Accelerator driven Subcritical core with Thorium as a new generation nuclear fuel. In this design core , A subcritical core coupled to an accelerator with proton beam (E p =1 GeV) is simulated by MCNPX code .Although the main purpose of ADS systems are transmutation and use MA (Minor Actinides) as a nuclear fuel but another use of these systems are use thorium fuel. This simulated core has two fuel assembly type : (Th-U) and (U-Pu) . Consequence , Neutronic parameters related to ADS core are calculated. It has shown that Thorium fuel is use able in this core and less nuclear waste ,Although Iran has not Thorium reserves but study on Thorium fuel cycle can open a new horizontal in use nuclear energy as a clean energy and without nuclear waste

  15. Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

    Energy Technology Data Exchange (ETDEWEB)

    Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

    1976-07-01

    Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

  16. Provision of reliable core cooling in vessel-type boiling reactors

    International Nuclear Information System (INIS)

    Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

    1987-01-01

    Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

  17. Seismic responses of N-Reactor core. Independent review of Phase II work

    International Nuclear Information System (INIS)

    Chen, J.C.; Lo, T.; Chinn, D.J.; Murray, R.C.; Johnson, J.J.; Maslenikov, O.R.

    1985-08-01

    Seismic response of the N-Reactor core was independently analyzed to validate the results of Impell's analysis. The analysis procedure consists of two major stages: linear soil-structure interaction (SSI) analysis of the overall N-Reactor structure complex and nonlinear dynamic analysis of the reactor core. In the SSI analysis, CLASSI computer codes were used to calculate the SSI response of the structures and to generate the input motions for the nonlinear reactor core analysis. In addition, the response was compared to the response from the SASSI analysis under review. The impact of foundation modeling techniques and the effect of soil stiffness variation on SSI response were also investigated. In the core analysis, a nonlinear dynamic analysis model was developed. The stiffness representation of the model was calculated through a finite element analysis of several local core geometries. Finite element analyses were also used to study the block to block interaction characteristics. Using this nonlinear dynamic model along with the basemat time histories generated from CLASSI and SASSI, several dynamic analyses of the core were performed. A series of sensitivity studies was performed to investigate the discretization of the core, the effect of vertical acceleration, the effect of basemat rocking, and modeling assumptions. In general, our independent analysis of core response validates the order of magnitude of the displacement calculated by Impell. 11 refs., 110 figs., 12 tabs

  18. Effective height of the core of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam); Martin, D P; Yip, F G [High Institute of Nuclear Sciences and Technology (Cuba)

    1994-10-01

    Measurements of thermal neutron relative distributions in axial direction at different positions in the reactor core and for various control rod configurations have been carried out, and axial buckling and effective height of the core deduced. (author). 4 refs., 3 figs., 1 tab.

  19. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure

  20. Isostructural solid-solid phase transition in monolayers of soft core-shell particles at fluid interfaces: structure and mechanics.

    Science.gov (United States)

    Rey, Marcel; Fernández-Rodríguez, Miguel Ángel; Steinacher, Mathias; Scheidegger, Laura; Geisel, Karen; Richtering, Walter; Squires, Todd M; Isa, Lucio

    2016-04-21

    We have studied the complete two-dimensional phase diagram of a core-shell microgel-laden fluid interface by synchronizing its compression with the deposition of the interfacial monolayer. Applying a new protocol, different positions on the substrate correspond to different values of the monolayer surface pressure and specific area. Analyzing the microstructure of the deposited monolayers, we discovered an isostructural solid-solid phase transition between two crystalline phases with the same hexagonal symmetry, but with two different lattice constants. The two phases corresponded to shell-shell and core-core inter-particle contacts, respectively; with increasing surface pressure the former mechanically failed enabling the particle cores to come into contact. In the phase-transition region, clusters of particles in core-core contacts nucleate, melting the surrounding shell-shell crystal, until the whole monolayer moves into the second phase. We furthermore measured the interfacial rheology of the monolayers as a function of the surface pressure using an interfacial microdisk rheometer. The interfaces always showed a strong elastic response, with a dip in the shear elastic modulus in correspondence with the melting of the shell-shell phase, followed by a steep increase upon the formation of a percolating network of the core-core contacts. These results demonstrate that the core-shell nature of the particles leads to a rich mechanical and structural behavior that can be externally tuned by compressing the interface, indicating new routes for applications, e.g. in surface patterning or emulsion stabilization.

  1. Optimized core loading sequence for Ukraine WWER-1000 reactors

    International Nuclear Information System (INIS)

    Dye, M.; Shah, H.

    2015-01-01

    Fuel Assemblies (WFAs) experienced mechanical damage of the grids during loading at both South Ukraine 2 (SU2) and South Ukraine 3 (SU3). The grids were damaged due to high lateral loads exceeding their strength limit. The high lateral loads were caused by a combination of distortion and stiffness of the mixed core fuel assemblies and significant fuel assembly-to-fuel assembly interaction combined with the core loading sequence being used. To prevent damage of the WFA grids during core loading, Westinghouse has developed a loading sequence technique and loading aides (smooth sided dummies and top nozzle loading guides) designed to minimize fuel assembly-to-fuel assembly interaction while maximizing the potential for successful loading (i.e., no fuel assembly damage and minimized loading time). The loading sequence technique accounts for cycle-specific core loading patterns and is based on previous Westinghouse WWER core loading experience and fundamental principles. The loading aids are developed to “open-up” the target core location or to provide guidance into a target core location. The Westinghouse optimized core loading sequence and smooth sided dummies were utilized during the successful loading of SU3 Cycle 25 mixed core in March 2015, with no instances of fuel assembly damage and yet still provided considerable time savings relative to the 2012 and 2013 SU3 reload campaigns. (authors)

  2. Modelling of reactor control and protection systems in the core simulator program GARLIC

    International Nuclear Information System (INIS)

    Beraha, D.; Lupas, O.; Ploegert, K.

    1984-01-01

    For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

  3. Study of Fuel Rods Axial Enrichment Distribution Effect on the Neutronic Parameters of the Reactor Core

    International Nuclear Information System (INIS)

    Pazirandeh, A.; Nasiri, S. H.

    2012-01-01

    Optimization of the fuel burn up is an important issue in nuclear reactor fuel management and technology. Radial enrichment distribution in the reactor core is a conventional method and axial enrichment is constant along the fuel rod. In this article, the effects of axial enrichment distribution variation on neutronic parameters of PWR core are studied. The axial length of the core is divided into ten sections, considering axial enrichment variation and leaving the existing radial enrichment distribution intact. This study shows that the radial and axial power peaking factors are decreased as compared with the typical conventional core. In addition, the first core lifetime lasts 30 days longer than normal PWR core. Moreover, at the same time boric acid density is 0.2 g/kg at the beginning of the cycle. The flux shape is also flat at the beginning of the cycle for the proposed configuration of the axially enrichment distribution.

  4. Reactor physics data for safety analysis of CANFLEX-NU CANDU-6 core

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun

    2001-08-01

    This report contains the reactor physics data for safety analysis of CANFLEX-NU fuel CANDU-6 core. First, the physics parameters for time-average core have been described, which include the channel power and maximum bundle power map, channel axial power shape and bundle burnup. And, next the data for fuel performance such as relative ring power distribution and bundle burnup conversion ratio are represented. The transition core data from 0 to 900 full power day are represented by 100 full power day interval. Also, the data for reactivity devices of time-average core and 300 full power day of transition core are given.

  5. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    Kotas, J.F.; Stroh, K.R.

    1983-01-01

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  6. An Idea of 20% test of the Initial Core Reactor Physics

    International Nuclear Information System (INIS)

    Roh, Kyung Ho; Yang, Sung Tae; Jung, Ji Eun

    2012-01-01

    Many tests have been performed on the OPR1000 and APR1400 before commercial operation. Some of these tests were performed at reactor power levels of 20% and 50%. The CPC (Core Protection Calculator) power distribution test is one of these tests. It is performed to assure the reliability of the Core Protection Calculator System (CPCS). Through this test, SAM1 is calculated using the snapshots2. The test takes about nine hours at a reactor power level of 20% and about thirty hours at a reactor power level of 50%. SAM used at each reactor power level is as follows: 1. Reactor power of 0% ∼ 20%: Designed SAM (DSAM) 2. Reactor power of 20% ∼ 50%: SAM calculated (C-SAM) at a reactor power of 20% 3. Reactor power 50% ∼ End of Cycle : SAM calculated at a reactor power of 50% As mentioned earlier, SAM is calculated and punched into CPC to assure the reliability of CPCS. Therefore, CPC is operated having penalties with D-SAM until3 reaching a reactor power of 20%. That is, the penalty of CPC will be removed when SAM is calculated and punched into the CPC at a reactor power of 20%. But these penalties are considered to be removed after a reactor power of 50% test in order to maintain the conservatism of the CPC. This is done because the final values calculated using C-SAM, in contrast to those calculated using SAM, a reactor power of 50%, are not correct. This paper began from an idea, 'If so, what would happen if we removed the CPC power distribution test at a reactor power of 20%?'

  7. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  8. A procedure for searching the equilibrium core of a research reactor

    International Nuclear Information System (INIS)

    Bakri Arbie; Liem Peng Hong; Prayoto

    1996-01-01

    A procedure for searching the equilibrium core of a research reactor has been proposed. The searching procedure is based on the relaxation method and has been implemented in Batan-EQUIL-2D in-core fuel management code. The few-group neutron diffusion theory in 2-D, X-Y, and R-Z reactor geometries are adopted as the framework of the code. The successful applicability of the procedure for obtaining the new RSG-GAS (MPR-30) silicide equilibrium core was shown. (author)

  9. INCA: method of analyzing in-core detector data in power reactors

    International Nuclear Information System (INIS)

    Ober, T.G.; Terney, W.B.; Marks, G.H.

    1975-04-01

    A method (INCA) is described by which signals from fixed in-core detectors are related to estimates of the three dimensional power distribution in an operating reactor core and to the maximum linear heat rate in the core. A description of the large library of data accompanying the method is provided. A detailed examination of the analytical verifications performed using the method is presented, and a summary of the uncertainty associated with the method is given. The uncertainty assigned to the maximum linear heat rate inferred by the method from operating reactor data is found to be 5.8 percent at a 95/95 confidence level. (U.S.)

  10. On the chemical constitution of a molten oxide core of a fast breeder reactor

    International Nuclear Information System (INIS)

    Hodkin, D.J.; Potter, P.E.

    1980-01-01

    A knowledge of the chemical constitution of a molten oxide fast reactor core is of great importance in the assessment of heat transfer from a cooling molten pool of debris and in the selection of materials for the construction of sacrificial beds for core containment. In this paper we describe some thermodynamic assessments of the likely chemical constitution of a molten oxide core, and then support our assessments by experimental observations

  11. Status of reactor core design code system in COSINE code package

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Liu, Z.

    2014-01-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  12. Status of reactor core design code system in COSINE code package

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

    2014-07-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  13. Design factors affecting dynamic behaviour of fast reactor cores. UK review paper

    Energy Technology Data Exchange (ETDEWEB)

    Brindley, K W [National Nuclear Corporation Ltd., Risley, Warrington (United Kingdom); Perks, M A [United Kingdom Atomic Energy Authority, Risley, Warrington (United Kingdom)

    1982-01-01

    This paper summarises the consideration that has been given in the UK to the following factors that affect the dynamic behaviour of fast reactor cores: fuel design - Pu/u homogeneity, fuel expansion, fuel-clad gaps, uranium fraction. Structural response - CR supports, diagrid, sub-assembly bowing sodium expansion coefficients - low void cores including heterogenous cores. Calculational methods and models are outlined and some experimental results are discussed. (author)

  14. Development of in-core measurements in the reactor KS-150

    International Nuclear Information System (INIS)

    Rana, S.B.

    1977-01-01

    Mapping of the neutron flux density distribution and of the neutron fluence distribution in the KS-150 reactor core was carried out using an in-core measuring system. The system allows the in-service monitoring of important operating properties of the reactor core and fuel elements and consists of a mapping fuel element assembly with built-in SPN detectors, of transmission paths and a computer facility. The measurement of the neutron flux, neutron fluence and temperature fields in the reactor core was carried out during the power start-up of the reactor using self-powered DPZ-1 detectors. The obtained data are given and the axial distribution of neutron flux is graphically represented for different values of burnup at the same configuration of regulating rods, as is the axial distribution of neutron fluence for different configurations of the regulating rods during operation, and the in-service neutron fluence distribution. The maximal fuel temperature of 500.2 degC was found at a distance of 291.2 cm from the upper boundary of the reactor core, at a neutron flux of 1.46x10 14 n/cm 2 s. In comparison with other methods, this method proved easy and quick, the results reliable, reactivity perturbance negligible and the fuel element cost increase a negligible 4%. Neutron flux mapping using in-core self-powered detectors will be performed on a wider scale. (J.P./J.O.)

  15. Radiological control aspects of the fabrication of the Light Water Breeder Reactor core (LWBR Development Program)

    International Nuclear Information System (INIS)

    Schultz, B.G.

    1979-05-01

    A description is presented of the radiological control aspects of the fabrication of the Light Water Breeder Reactor (LWBR) core. Included are the radiological control criteria applied for the design and use of fabrication facilities, the controls and limits imposed to minimize radiaion exposure to personnel, and an evaluation of the applied radiological program in meeting the program objectives. The goal of the LWBR program is to develop the technology to breed in light water reactors so that nuclear fuel may be used significantly more efficiently in these reactors. This technology is being developed by designing and fabricating a breeder reactor core, utilizing thoria (ThO 2 ) and binary thoria--urania (ThO 2 - 233 UO 2 ) fuel, to be operated in the existing pressurized water reactor plant owned by the Department of Energy at Shippingport, Pennsylvania

  16. Electron momentum spectroscopy of the core state of solid carbon

    International Nuclear Information System (INIS)

    Caprari, R.S.; Clark, S.A.C.; McCarthy, I.E.; Storer, P.J.; Vos, M.; Weigold, E.

    1994-08-01

    Electron momentum spectroscopy (binary encounter (e,2e)) experimental results are presented for the core state of an amorphous carbon allotrope. The (e,2e) cross section has two identifiable regions. One is a narrow energy width 'core band peak' that does not disperse with momentum. At higher binding energies there is an energy diffuse 'multiple scattering continuum', which is a consequence of (e,2e) collisions with core electrons that are accompanied by inelastic scattering of one or more of the incoming or outgoing electrons. Comparisons of experimental momentum distributions with the Hartree-Fock atomic carbon ls orbital are presented for both regions. 16 refs., 4 figs

  17. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    International Nuclear Information System (INIS)

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

  18. A system for obtaining an optimized pre design of nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.

    1989-01-01

    This work proposes a method for obtaing a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one-energy-group, unidimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, refletor thickness, enrichement and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for futures works. (author) [pt

  19. A system to obtain an optimized first design of a nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.

    1988-01-01

    This work proposes a method for obtaining a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one energy-group, one-dimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, reflector thickness, enrichment and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for future works. (autor)

  20. Development of cutting technique of reactor core internals by CO laser

    International Nuclear Information System (INIS)

    Takano, G.; Beppu, S.; Matsumoto, O.; Sakamoto, N.; Onozawa, T.; Sugihara, M.; Miya, K.

    1995-01-01

    The CO laser is superior in the absorption characteristic to materials to the CO 2 laser due to its shorter wavelength. In consideration of this characteristic Nuclear Power Engineering Corporation is studying this applicability sponsored by the Ministry of International Trade Industry of Japan to cutting of reactor core internals of commercial nuclear power plant. In decommissioning of reactor core internals it is necessary to cut stainless steel plates of 305 mm thick. The authors cut stainless steel plates of up to 310mm thick in air and those of up to 150 mm thick underwater with a 20kW class laser. Further, models simulating key structural elements of PWR core internals were cut and secondary products to clarify the applicability of the CO laser cutting to reactor core internals were evaluated. (author)

  1. Preliminary design of a borax internal core-catcher for a gas cooled fast reactor

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Schumacher, G.

    1976-09-01

    Preliminary thermal calculations show that a core-catcher appears to be feasible, which is able to cope with the complete meltdown of the core and blankets of a 1,000 MWe GCFR. This core-catcher is based on borax (Na 2 B 4 O 7 ) as dissolving material of the oxide fuel and of the fission products occuring in oxide form. The borax is contained in steel boxes forming a 2.1 meter thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel, just underneath the reactor core. The fission products are dispersed in the pool formed by the liquid borax. The heat power density in the pool is conveniently reduced and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system. (orig.) [de

  2. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

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

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

    Science.gov (United States)

    Putre, H. A.

    1973-01-01

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

  4. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  6. Tailored Core Shell Cathode Powders for Solid Oxide Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Swartz, Scott [NexTech Materials, Ltd.,Lewis Center, OH (United States)

    2015-03-23

    In this Phase I SBIR project, a “core-shell” composite cathode approach was evaluated for improving SOFC performance and reducing degradation of lanthanum strontium cobalt ferrite (LSCF) cathode materials, following previous successful demonstrations of infiltration approaches for achieving the same goals. The intent was to establish core-shell cathode powders that enabled high performance to be obtained with “drop-in” process capability for SOFC manufacturing (i.e., rather than adding an infiltration step to the SOFC manufacturing process). Milling, precipitation and hetero-coagulation methods were evaluated for making core-shell composite cathode powders comprised of coarse LSCF “core” particles and nanoscale “shell” particles of lanthanum strontium manganite (LSM) or praseodymium strontium manganite (PSM). Precipitation and hetero-coagulation methods were successful for obtaining the targeted core-shell morphology, although perfect coverage of the LSCF core particles by the LSM and PSM particles was not obtained. Electrochemical characterization of core-shell cathode powders and conventional (baseline) cathode powders was performed via electrochemical impedance spectroscopy (EIS) half-cell measurements and single-cell SOFC testing. Reliable EIS testing methods were established, which enabled comparative area-specific resistance measurements to be obtained. A single-cell SOFC testing approach also was established that enabled cathode resistance to be separated from overall cell resistance, and for cathode degradation to be separated from overall cell degradation. The results of these EIS and SOFC tests conclusively determined that the core-shell cathode powders resulted in significant lowering of performance, compared to the baseline cathodes. Based on the results of this project, it was concluded that the core-shell cathode approach did not warrant further investigation.

  7. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  8. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  9. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  10. Sources of gamma radiation in a reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Roos, Matts

    1959-05-15

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

  11. Emergency core cooling system for a fast reactor

    International Nuclear Information System (INIS)

    Johnson, H.G.; Madsen, R.N.

    1976-01-01

    The main heat transport system for a liquid-metal-cooled nuclear reactor is constructed with elevated piping and guard vessels or pipes around all components of the system below the elevation of the elevated piping so the head developed by the pumps at emergency motor speed will be unsufficient to lift the liquid-metal-coolant over the top of the guard tanks or pipes or out of the elevated piping in the event of a loss-of-coolant accident. In addition, inlet downcomers to the reactor vessel are contained within guard standpipes having a clearance volume as small as practicable. 4 claims, 2 drawing figures

  12. Core construction in a pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  13. The development of ex-core neutron flux monitoring system for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S

    2004-12-01

    Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

  14. The development of ex-core neutron flux monitoring system for integral reactor

    International Nuclear Information System (INIS)

    Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S.

    2004-12-01

    Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying the ex-core

  15. Method of core thermodynamic reliability determination in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ackermann, G.; Horche, W. (Ingenieurhochschule Zittau (German Democratic Republic). Sektion Kraftwerksanlagenbau und Energieumwandlung)

    1983-01-01

    A statistical model appropriate to determine the thermodynamic reliability and the power-limiting parameter of PWR cores is described for cases of accidental transients. The model is compared with the hot channel model hitherto applied.

  16. Method of core thermodynamic reliability determination in pressurized water reactors

    International Nuclear Information System (INIS)

    Ackermann, G.; Horche, W.

    1983-01-01

    A statistical model appropriate to determine the thermodynamic reliability and the power-limiting parameter of PWR cores is described for cases of accidental transients. The model is compared with the hot channel model hitherto applied. (author)

  17. Operation and Licensing of Mixed Cores in Water Cooled Reactors

    International Nuclear Information System (INIS)

    2013-11-01

    Nuclear fuel is a highly complex material that is subject to continuous development and is produced by a range of manufacturers. During operation of a nuclear power plant, the nuclear fuel is subject to extreme conditions of temperature, corroding environment and irradiation, and many different designs of fuel have been manufactured with differing fuel materials, cladding materials and assembly structure to ensure these conditions. The core of an operating power plant can contain hundreds of fuel assemblies, and where there is more than a single design of a fuel assembly in the core, whether through a change of fuel vendor, introduction of an improved design or for some other reason, the core is described as a mixed core. The task of ensuring that the different assembly types do not interact in a harmful manner, causing, for example, differing flow resistance resulting in under cooling, is an important part of ensuring nuclear safety. This report has compiled the latest information on the operational experience of mixed cores and the tools and techniques that are used to analyse the core operation and demonstrate that there are no safety related problems with its operation. This publication is a result of a technical meeting in 2011 and a series of consultants meetings

  18. A New In-core Production Method of Co-60 in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lyu, Jinqi; Kim, Woosong; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of); Park, Younwon [BEES Inc, Daejeon (Korea, Republic of)

    2016-05-15

    This study introduces an innovative method for Co-60 production in the CANDU6 core. In this new scheme, the central fuel element is replaced by a Co-59 target and Co-60 is obtained after the fuel bundle is discharged. It has been shown that the new method can produce significantly higher amount of Co-60 than the conventional Co production method in CANDU6 reactors without compromising the fuel burnup by removing some (<50%) of the adjuster rods in the whole core. The coolant void reactivity is noticeably reduced when a Co-59 target is loaded into the central pin of the fuel bundle. Meanwhile, the peak power in a fuel bundle is just a little higher due to the central Co-59 target than in conventional CANDU6 fuel design. The basic technology for Co-60 producing was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) in 1946 and the same technology was adapted and applied in CANDU6 power reactors. The standard CANDU6 reactor has 21 adjuster rods which are fully inserted into the core during normal operation. The stainless steel adjuster rods are replaced with neutronically-equivalent Co-59 adjusters to produce Co-60. Nowadays, the roles of the adjuster rods are rather vague since nuclear reactors cannot be quickly restarted after a sudden reactor trip due to more stringent regulations. In some Canadian CANDU6 reactors, some or all the adjuster rods are removed from the core to maximize the uranium utilization.

  19. 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)

  20. Transient bowing of core assemblies in advanced liquid metal fast reactors

    International Nuclear Information System (INIS)

    Kamal, S.A.; Orechwa, Y.

    1986-01-01

    Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

  1. The reactor core configuration and important systems related to physics tests of Daya Bay NPP

    International Nuclear Information System (INIS)

    Tao Shaoping

    1995-06-01

    A brief introduction to reactor core configuration and important systems related to physics tests of Daya Bay NPP is given. These systems involve the reactor core system (COR), the full length rod control system (RGL), the in-core instrumentation system (RIC), the out-of-core nuclear instrumentation system (RPN), and the LOCA surveillance system (LSS), the centralized data processing system (KIT) and the test data acquisition system (KDO). In addition, that the adjustment and evaluation of boron concentration related to other systems, for example the reactor coolant system (RCP), the chemical and volume control system (RCV), the reactor boron and water makeup system (REA), the nuclear sampling system (REN) and the reactor control system (RRC), etc. is also described. Analysis of these systems helps not only to familiarize their functions and acquires a deepen understanding for the principle procedure, points for attention and technical key of the core physics tests, but also to further analyze the test results. (3 refs., 11 figs., 1 tab.)

  2. Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-11-01

    The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

  3. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    International Nuclear Information System (INIS)

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR P C package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change

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

    International Nuclear Information System (INIS)

    Markovic, H.

    1962-07-01

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

  5. HORECA. Hoger onderwijs reactor elementary core analysis system. User's manual

    International Nuclear Information System (INIS)

    Battum, E. van; Serov, I.V.

    1993-07-01

    HORECA is developed at IRI Delft for quick analysis of power distribution, burnup and safety for the HOR. It can be used for the manual search of a better loading of the reactor. HORECA is based on the Penn State Fuel Management Package and uses the MCRAC code included in this package as a calculation engine. (orig./HP)

  6. Transient neutrons flux behaviour in a spherical reactor core

    International Nuclear Information System (INIS)

    Souza, A.W.A. de.

    1978-11-01

    This work studies the transient neutron flux in a fast reactor of spherical geometry. The burning of U 235 nuclei is equated and two kinds of reflector were studied. The numeric solutions are then compared with the results for those reflectors. (author) [pt

  7. An approach to model reactor core nodalization for deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  8. An approach to model reactor core nodalization for deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

    2016-01-22

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  9. An approach to model reactor core nodalization for deterministic safety analysis

    International Nuclear Information System (INIS)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH 1.6 , stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D ® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M

  10. Modelling of non-catalytic reactors in a gas-solid trickle flow reactor: Dry, regenerative flue gas desulphurization using a silica-supported copper oxide sorbent

    NARCIS (Netherlands)

    Kiel, J.H.A.; Kiel, J.H.A.; Prins, W.; van Swaaij, Willibrordus Petrus Maria

    1992-01-01

    A one-dimensional, two-phase dispersed plug flow model has been developed to describe the steady-state performance of a relatively new type of reactor, the gas-solid trickle flow reactor (GSTFR). In this reactor, an upward-flowing gas phase is contacted with as downward-flowing dilute solids phase

  11. Effect of core burnup on the dynamic behavior of fast reactors

    International Nuclear Information System (INIS)

    Ilberg, D.; Saphier, D.; Yiftah, S.

    1977-01-01

    Performance of a dynamic analysis, taking burnup changes into account, requires fission-product nuclear data of relatively small uncertainty, suitable burnup calculation models, and dynamic computer programs. These were prepared and used with the following results: (1) Significant changes in static and dynamic parameters were observed when investigating the effect of burnup. These changes were found to be larger than differences introduced by the uncertainty of the fission-product nuclear data. (2) A one-dimensional burnup computer program was prepared. It was found that a burnup model based on the generalized radioactive decay scheme is suitable for accurate fast reactor calculations. (3) Space-time dynamic calculations of fast reactors having different burnup levels were performed. The stability difference between ''clean'' and high burnup cores is greater when local rather than uniform perturbations are inserted along the entire core length. The magnitude by which the ''end-of-life'' core increases the transient excursion over that of the clean core depends on the particular region in which the perturbation is inserted. The end-of-life core will magnify the transient excursion more than the clean core whenever the perturbation is inserted into a region having a higher adjoint flux level than that of the clean core. However, when a reactor safety system operates successfully, the difference in the temperature transient of the clean and end-of-life cores will be relatively small. It is suggested that only the analysis of large local perturbations be performed for end-of-life cores as well as for clean cores in the safety evaluation of fast reactors

  12. Upper-bound fission product release assessment for large break LOCA in CANFLEX bundle reactor core

    International Nuclear Information System (INIS)

    Oh, Duk Ju; Lee, Kang Moon

    1996-07-01

    Quarter-core gap inventory assessment for CANDU-6 reactor core loaded with CANFLEX fuel bundles has been performed as one of the licensing safety analyses required for 24 natural uranium CANFLEX bundle irradiation in CANDU-6 reactor. The quarter-core gap inventory for the CANFLEX bundle core is 5 - 10 times lower than that for the standard bundle core, depending on the half-life of the isotope. The lower gap inventory of the CANFLEX bundle core is attributed to the lower linear power of the CANFLEX bundle compared with the standard bundle. However, the whole core total inventories for both the CANFLEX and standard bundle cores are nearly the same. The 6 - 8 times lower upper-bound fission product releases of the CANFLEX bundle core for large break LOCA than those of the standard bundle core imply that the loading of 24 natural uranium CANFLEX bundles would improve the predicted consequences of the postulated accident described in the Wolsung 2 safety report. 2 tabs., 6 figs., 3 refs. (Author)

  13. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  14. Mechanical design of core components for a high performance light water reactor with a three pass core

    International Nuclear Information System (INIS)

    Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

    2007-01-01

    Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

  15. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core

    International Nuclear Information System (INIS)

    Lucatero, M.A.; Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

    2006-01-01

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, rΘ, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  16. A study on reactor core failure thresholds to safety operation of LMFBR

    International Nuclear Information System (INIS)

    Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo

    2006-01-01

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  17. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

    2017-07-11

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

  18. Assessment of the residual time to rupture of fuel pins after reactor core disturbances using the Lebensanteil rule

    International Nuclear Information System (INIS)

    Schaefer, L.; Wassilew, C.

    1992-01-01

    An important aspect of disturbances in the reactor core is the way in which they affect the service life of fuel rod cladding tubes. This factor also determines whether and how long the reactor core can be continued in operation, i.e., matters of safety and economy are involved. Potential disturbances of the reactor core affect the fuel rod cladding tubes as increases in temperature and, sometimes, as mechanical stresess for limited periods of time. As thermomechanical stresses acting on a cladding tube also give rise to creep events which may limit the service life of fuel elements, it is important to know how much creep life or time to rupture is consumed in the course of a core disturbance, and what the residual life is. For this purpose, the times to rupture before and during the accident must be added up and the balance calculated. As a rule of computation, the Lebensanteil rule is used in its form expressing the time to rupture of creeping solids. The simulation of accidents with unirradiated cladding tubes revealed a drastic decrease of the residual time to rupture in those cases in which the cladding material had recrystallized. On the other hand, because of its structural stability, irradiated material turned out to be almost insensitive even under extremely severe accident conditions. The materials data so far available are sufficient for useful estimates. Presuming one of the damage accumulating processes of the creeping cladding material is predominant, there are no further validity limiting ranges concerning kind of accident, loading condition, cladding material and so on. (orig.)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  20. Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2011-01-01

    The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)