WorldWideScience

Sample records for reactor core containment

  1. 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

  2. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

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

  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. 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.)

  5. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

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

  6. 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.)

  7. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  8. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  9. A reactor core/containment status evaluation flowchart for determining protective actions in emergencies

    International Nuclear Information System (INIS)

    Glissman, M.A.

    1988-01-01

    In the event of an emergency at a power reactor station, there might not be adequate time or sufficient data to fully assess radiological implications and make protective action recommendations based on projected population exposures. Thus, decision-making guidance is needed that is based on readily available plant indicators, not just on time-consuming dose calculations. In the United States, this guidance must be compatible with the recommended by the Nuclear Regulatory Commission and the Environmental Protection Agency, and it must include predetermined, measurable, site-specific parameters for assessing conditions in the reactor core and containment. The preparation of this real time guidance calls for the selection of suitable parameters and the determination of the values for these parameters that will correspond to different levels of protective action. This process is illustrated in this paper by selecting parameters and determining appropriate values for constructing a Core/Containment Status Evaluation Flowchart for an example power plant

  10. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  11. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  12. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  13. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  14. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  15. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  16. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  17. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  18. 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.)

  19. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  20. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  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. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  3. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

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

  4. Reactor container facility

    International Nuclear Information System (INIS)

    Saito, Takashi; Nagasaka, Hideo.

    1990-01-01

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

  5. 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

  6. Reactor container structure

    International Nuclear Information System (INIS)

    Sato, Yoshimi; Fukuda, Yoshio.

    1993-01-01

    A main container of an FBR type reactor using liquid sodium as coolants is attached to a roof slug. The main container contains, as coolants, lower temperature sodium, and high temperature sodium above a reactor core and a partitioning plate. The main container has a structure comprising only longitudinal welded joints in parallel with axial direction in the vicinity of the liquid surface of high temperature sodium where a temperature gradient is steep and great thermal stresses are caused without disposing lateral welded joints in perpendicular to axial direction. Only the longitudinal welded joints having a great fatigue strength are thus disposed in the vicinity of the liquid surface of the high temperature sodium where axial thermal stresses are caused. This can improve reliability of strength at the welded portions of the main container against repeating thermal stresses caused in vicinity of the liquid surface of the main container from a view point of welding method. (I.N.)

  7. 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

  8. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  9. 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.)

  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 container spray device

    International Nuclear Information System (INIS)

    Yanai, Ryoichi.

    1980-01-01

    Purpose: To enable decrease in the heat and the concentration of radioactive iodine released from the reactor vessel into the reactor container in the spray device of BWR type reactors. Constitution: A plurality of water receiving trays are disposed below the spray nozzle in the dry well and communicated to a pressure suppression chamber by way of drain pipeways passing through a diaphragm floor. When the recycling system is ruptured and coolants in the reactor vessel and radioactive iodine in the reactor core are released into the dry well, spray water is discharged from the spray nozzle to eliminate the heat and the radioactive iodine in the dry well. In this case, the receiving trays collect the portions of spray water whose absorption power for the heat and radioactive iodine is nearly saturated and falls them into the pool water of the pressure suppression chamber. Consequently, other portions of the spray water that still possess absorption power can be jetted with no hindrance, to increase the efficiency for the removal of the heat and iodine of the spray droplets. (Horiuchi, T.)

  13. 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.)

  14. 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.)

  15. 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.

  16. Fission reactor container

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1991-01-01

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

  17. 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.)

  18. Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995

    International Nuclear Information System (INIS)

    Sforza, P.M.; Cresci, R.J.

    1997-01-01

    Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical

  19. 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.)

  20. 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.)

  1. 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.)

  2. 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.)

  3. Safety system for reactor container

    International Nuclear Information System (INIS)

    Shimizu, Miwako; Seki, Osamu; Mano, Takio.

    1995-01-01

    A slanted structure is formed below a reactor core where there is a possibility that molten reactor core materials are dropped, and above a water level of a pool which is formed by coolants flown from a reactor recycling system and accumulated on the inner bottom of the reactor container, to prevent molten fuels from dropping at once in the form of a large amount of lump. The molten materials are provisionally received on the structure, gradually formed into small pieces and then dropped. Further, the molten materials are dropped and received provisionally on a group of coolant-flowing pipelines below the structure, to lower the temperature of the molten materials, and then the reactor core molten materials are gradually formed into small pieces and dropped into the pool water. Since they are not dropped directly into the pool water but dropped gradually into the pool water as small droplets, occurrence of steam explosion can be reduced. The occurrence of steam explosion due to dropped molten reactor core material and pool water is suppressed, and the molten materials are kept in the pool water, thereby enabling to maintain the integrity of the reactor container more effectively. (N.H.)

  4. 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

  5. Nuclear reactor container

    International Nuclear Information System (INIS)

    Ishiyama, Takenori.

    1989-01-01

    This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

  6. Nuclear reactor container

    International Nuclear Information System (INIS)

    Shioiri, Akio.

    1992-01-01

    In a nuclear reactor container, a vent tube communication port is disposed to a pressure suppression pool at a position higher than the pool water therein for communication with an upper dry well, and the upper end opening of a dry well communication pipe is disposed at a position higher than the communication port. When condensate return pipeline is ruptured in the upper dry well, water in a water source pool is injected to the pressure vessel and partially discharged out of the ruptured port and a depressurization valve connected to the pressure vessel to the inside of the upper dry well. The discharged water stays in the upper dry well and, when the water level reaches the height of the vent tube communication port, it flows into the pressure suppression pool. Even in a state that the entire amount of water in the water source pool is supplied, since water does not reach the upper opening port of the dry well communication pipe, water does not flow into a lower dry well. Accordingly, the motor of a control rod drives disposed in the lower dry well can be prevented from submerging. The reactor core can be cooled more reliably, to improve the reliability of the pressure suppression function. (N.H.)

  7. 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.)

  8. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-02-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

  9. 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.)

  10. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Arai, Kenji.

    1996-01-01

    Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

  11. 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.)

  12. 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.)

  13. 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

  14. 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.)

  15. 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)

  16. Nuclear reactor container

    International Nuclear Information System (INIS)

    Hosaka, Seiichi.

    1988-01-01

    Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

  17. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

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

  18. Reactor containing facility

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1992-01-01

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

  19. Nuclear reactor container

    International Nuclear Information System (INIS)

    Yamaki, Rika; Kawabe, Ryuhei.

    1989-01-01

    A venturi scrubber is connected to a nuclear reactor container. Gases containing radioactive aerosols in the container are introduced into the venturi scrubber in the form of a high speed stream under the pressure of the container. The radioactive aerosols are captured by inertia collision due to the velocity difference between the high speed gas stream and water droplets. In the case of the present invention, since the high pressure of the reactor container generated upon accident is utilized, compressor, etc. is no more required, thereby enabling to reduce the size of the aerosol removing device. Further, since no external power is used, the radioactive aerosols can be removed with no starting failure upon accidents. (T.M.)

  20. CONCRETE REACTOR CONTAINMENT

    Energy Technology Data Exchange (ETDEWEB)

    Lumb, Ralph F.; Hall, William F.; Fruchtbaum, Jacob

    1963-06-15

    The results of various leak-rate tests demonstrate the practicality of concrete as primary containment for the maximum credible accident for a research reactor employing plate-type fuel and having a power in excess of one megawatt. Leak-test time was shortened substantially by measuring the relaxation time for overpressure decay, which is a function of leak rate. (auth)

  1. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  2. 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.)

  3. 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

  4. 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

  5. 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)

  6. Nuclear reactor container

    International Nuclear Information System (INIS)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Murase, Michio; Fujii, Tadashi; Susuki, Akira.

    1994-01-01

    A wet well space above a pressure suppression pool is divided into a first wet well on the side in contact with the pressure suppression pool and a second wet well on the side not in contact with the pool. Cooling water is contained in the second wet well and it is in communication with the first wet well by pipelines. Since steams flown into the second well are condensed in the cooling water, they continuously transfer from the first wet well to the second wet well, thereby capable of eliminating the effects of incondensible gases in the first wet well. With such procedures, the effect of the incondensible gases can be eliminated even without cooling from the outside of the reactor. Heat accumulation can be increased in a container of any material, so that thermal load on cooling circuits for removing after-heat can be mitigated. (T.M.)

  7. 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.)

  8. 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

  9. 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)

  10. Nuclear reactor container

    International Nuclear Information System (INIS)

    Fukui, Tooru; Murase, Michio; Kataoka, Yoshiyuki; Hidaka, Masataka; Sumita, Isao; Tominaga, Kenji.

    1992-01-01

    In a nuclear reactor container, a chamber in communication with a wet well of a pressure suppression chamber is disposed and situated to such a position that the temperature is lower than a chamber containing pool water upon occurrence of loss of coolant accident. In addition, the inner surface of the pressure suppression chamber is constituted with steel walls in contact with pool water, and an outer circumferential pool is disposed at the outer circumferential surface thereof. Further, a circulation channel is disposed, and a water intake port is disposed at a position higher than an exit to the pool water, and a water discharge port is opened in the pool water at a position lower than the exit to the pool water. With such a constitution, the allowable temperature of the pressure suppression pool water can be elevated to a saturated steam temperature corresponding to the resistant pressure of the container, so that the temperature difference between the pressure suppression pool and the outer side thereof is increased by so much, to improve thermal radiation performance. Accordingly, it can be utilized as a pressure suppression means for a plant of greater power. Further, thermal conduction efficiency from the pool water region of the pressure suppression chamber to the outer circumferential pool water is improved, or thermal radiation area is enlarged due to the circulation channel, to improve the heat radiation performance. (N.H.)

  11. Nuclear reactor container

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1989-01-01

    Aerosol filters considered so far for nuclear reactor containers in conventional BWR type nuclear power plants make the facility larger and involve a risk of clogging. In view of the above, in the present invention, the diameter of a flow channel of gases entering from a bent pipe to a suppression pool is made smaller thereby decreasing the diameter of gas bubbles in the supperssional pool. Since this reduces the force of surface tension, the diameter of resulted gas bubbles is made remarkably smaller as compared with the case where the gases are released from the lower end of the bent pipe. Since the absorption velocity of bubble-entrained aerosols into water is in proportion to the square of the bubble diameter, the absorption efficiency can be increased remarkably by reducing the diameter of the gas bubbles. Accordingly, it is possible to improve the efficiency of eliminating radioactivity of released gases. (K.M.)

  12. Reactor containment vessel

    International Nuclear Information System (INIS)

    Ochiai, Kanehiro; Hayagumo, Sunao; Morikawa, Matsuo.

    1981-01-01

    Purpose: To safety and simplify the structure in a reactor containment vessel. Constitution: Steam flow channels with steam jetting ports communicating to coolants are provided between a communication channel and coolants in a pressure suppression chamber. Upon loss of coolant accidents, pressure in a dry well will increase, then force downwards water in an annulus portion and further flow out the water through steam jetting ports into a suppression pool. Thus, the steam flow channel is filled with steams or airs present in the dry well, which are released through the steam jetting ports into the pressure suppression chamber. Even though water is violently vibrated owing to the upward movement of air bubbles and condensation of steam bubbles, the annular portion and the steam jetting ports are filled with steams or the like, direct dynamic loads onto the structures such as communication channels can be avoided. (J.P.N.)

  13. Water injection device for reactor container

    International Nuclear Information System (INIS)

    Sakaki, Isao.

    1996-01-01

    A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

  14. 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.)

  15. Nuclear reactor container

    International Nuclear Information System (INIS)

    Moriyama, Takeo; Ochiai, Kanehiro; Niino, Tsuyoshi; Kodama, Toyokazu; Hirako, Shizuka.

    1988-01-01

    Purpose: To obtain structures suitable to a container structures for nuclear power plants used in those districts where earthquakes occur frequently, in which no local stresses are caused to the fundamental base portions and the workability for the fundamental structures is improved. Constitution: Basic stabilizers are attached to a nuclear reactor container (PCV) and a basic concrete recess for receiving a basic stabilizer is disposed in basic concretes. A top stabilizer is joined and fixed to a top stabilizer receiving plate at the inside of an outer shielding wall. On the other hand, a PCV top recess for conducting the load of PCV to the top stabilizer is attached to the top of the PCV. By disposing stabilizer structures allowing miner displacements at the two points, that is, the top and the lowermost portion of the PCV, no local stress concentrations can be generated to the extension on the axial direction of components due to the inner pressure of the PCV and to the horizontal load applied to the upper portion of the PCV upon earthquakes. (Yoshino, Y.)

  16. 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

  17. 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.)

  18. 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

  19. 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)

  20. 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.)

  1. 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.).

  2. 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

  3. 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.)

  4. Mobile nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Thompson, R.E.; Spurrier, F.R.; Jones, A.R.

    1978-01-01

    A containment vessel for use in mobile nuclear reactor installations is described. The containment vessel completely surrounds the entire primary system, and is located as close to the reactor primary system components as is possible in order to minimize weight. In addition to being designed to withstand a specified internal pressure, the containment vessel is also designed to maintain integrity as a containment vessel in case of a possible collision accident

  5. 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.)

  6. 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.)

  7. 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.)

  8. Emergency reactor container cooling facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Matsumoto, Tomoyuki.

    1992-01-01

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

  9. 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

  10. 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.)

  11. 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.)

  12. Heat removing device for reactor container

    International Nuclear Information System (INIS)

    Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

    1996-01-01

    A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

  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. Containment for low temperature district nuclear-heating reactor

    International Nuclear Information System (INIS)

    He Shuyan; Dong Duo

    1992-03-01

    Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

  15. Humidity control device in a reactor container

    International Nuclear Information System (INIS)

    Aizawa, Motohiro; Igarashi, Hiroo; Osumi, Katsumi; Kimura, Takashi.

    1986-01-01

    Purpose: To provide a device capable of maintaining the inside of a container under high humidity circumstantial conditions causing less atmospheric corrosions, in order to prevent injuries due to atmospheric corrosions to smaller diameter stainless steel pipeways in the reactor container. Constitution: Stress corrosion cracks (SCC) to the smaller diameter stainless steel pipeways are caused dependent on the relative humidity and it is effective as the countermeasure against SCC to maintain the relative humidity at a low level less than 30 % or high level greater than 60 %. Based on the above findings, a humidity control device is disposed so as to maintain the relative humidity for the atmosphere within a reactor core on a higher humidity region. The device is adapted such that recycling gas in the dry-well coolant circuit is passed through an orifice to atomize the water introduced from feedwater pipe and introduce into a reactor core or such that the recycling gases in the dry-well cooling circuit are bubbled into water to remove chlorine gas in the reactor container gas thereby increasing the humidity in the reactor container. (Kamimura, M.)

  16. 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.)

  17. APROS couplings from core to containment

    International Nuclear Information System (INIS)

    Puska, E.K.; Ylijoki, J.

    2005-01-01

    APROS simulation environment is able to describe the 1-D and 3-D neutronics of the reactor core. It is also able to describe the thermal hydraulics of the core and circuits either with 5- equation or 6-equation thermal hydraulics. It can also describe the plant automation and electrical systems, as well as the behaviour of the containment. The peculiar feature of APROS in comparison to other coupled systems is that all parts in the coupled system are described with the same code instead of coupling two or three separate codes together with information exchange between the separate codes. The most recent possibility is the coupled calculation of the process and the containment. The more traditional coupling, the coupling of the process containing both the process description and the automation description with more or less detailed description of the 3-D core either for safety analysis or real-time simulation purposes has been discussed in previous work. The paper presents and discusses the capabilities of the code in coupling the plant process and automation description with the plant containment description with two example transient cases. An improved boron concentration solution with second order upwind discretization has been recently included in APROS. An example on the increased accuracy acquired in the 3-D core model has been included. (authors)

  18. 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

  19. Upgrade of the Annular Core Pulse Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-07-01

    elements will occupy the three rings adjacent to the irradiation cavity and the U-ZrH elements will be placed in the four outer rings. Other modifications are anticipated for the reactor systems. A completely new control system will be installed, a larger pool cooling capability is required for a higher steady-state power, and the need for a containment system is anticipated in the event of a reactor accident. Another part of the fast reactor safety research program at Sandia Laboratories is the investigation of techniques to observe the motion of molten fuel and clad during and after a reactor pulse. Such a device may require changes to the grid configuration so that a radial row of fuel elements can be removed to allow observation of an experiment in the cavity. Core design considerations for a fuel motion detection system are a part of the upgrade project. (author)

  20. 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

  1. 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)

  2. 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

  3. The containment safety of the Dragon Reactor

    International Nuclear Information System (INIS)

    Cullington, G.R.

    1967-08-01

    The original design of the Dragon Reactor was based upon the assumption that fission product emitting fuel elements would be used, leading to two significant considerations. First, a highly active primary circuit would result in normal operation, and second, under accident conditions involving massive core damage and corrosion following a major pressure vessel failure, the bulk of the core burden of fission products would be released. The adoption of coated particle fuel able to retain fission products has changed significantly the philosophy behind the design of the containment. The new philosophy is described and its effect on operating principles is discussed. (UK)

  4. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Akiba, Miyuki.

    1996-01-01

    In a cooling device for a reactor container, a low pressure vessel is connected to an incondensible gas vent tube by way of an opening/closing valve. Upon occurrence of a loss of coolant accident, among steams and incondensible gases contained in the reactor container, steams are cooled and condensed in a heat exchanger. The incondensible gases are at first discharged from the heat exchanger to a suppression pool by way of the incondensible gas vent tube, but subsequently, they are stagnated in the incondensible gas vent tube to hinder heat exchanging and steam cooling and condensing effects in the heat exchanger thereby raising temperature and pressure in the reactor. However, if the opening/closing valve is opened when the incondensible gases are stagnated in the incondensible gas vent tube, since the incondensible gases stagnated in the heat exchanger are sucked and discharged to the low pressure vessel, the performance of the heat exchanger is maintained satisfactorily thereby enabling to suppress elevation of temperature and pressure in the reactor container. (N.H.)

  5. 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)

  6. 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

  7. Light and heavy water replacing system in reactor container

    International Nuclear Information System (INIS)

    Miyamoto, Keiji.

    1979-01-01

    Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

  8. 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.)

  9. Pressure releasing device for reactor container

    International Nuclear Information System (INIS)

    Takeda, Mika.

    1994-01-01

    In the present invention, dose rate to public caused by radioactive rare gases can be decreased. That is, a reactor container contains a reactor pressure vessel incorporating a reactor core. There are disposed a pressure releasing system for releasing the pressure in the reactor pressure vessel to the outside, and a burning device for burning gases released from the pressure releasing system. An exhaustion pipe is disposed to the pressure releasing system. A burning device is disposed to the exhaustion pipe. It is effective to dispose a ventilation port at a portion of the exhaustion pipe upstream of the burning device. In addition, the burning device may preferably be disposed in a multi-stage in the axial direction of the exhaustion pipe. With such procedures, hydrogen in gases discharged along with the release of the pressure in the container is burned. Buoyancy is caused to the exhaustion gases by heat energy upon burning. Since the exhaustion gases can reach a higher level by the buoyancy, the dose rate due to the rare gases can be reduced. (I.S.)

  10. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

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

  11. 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.)

  12. 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)

  13. 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.)

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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.)

  5. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  6. 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

  7. 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

  8. 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)

  9. Cooling facility for reactor container

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro

    1996-05-31

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

  10. Cooling facility for reactor container

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  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. 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)

  13. 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.)

  14. 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

  15. 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.)

  16. Elements of thought on corium containment strategy in reactor vessel

    International Nuclear Information System (INIS)

    2015-01-01

    As accidents with core fusion are taken into account for the design of third-generation nuclear reactors, this brief document presents the corium containment strategy for a reactor vessel, its limitations, as well as research programs undertaken by the IRSN in this field. The report describes the controlled management of a severe accident, the major objective being to minimise releases in the environment, that which requires to maintain the reactor containment enclosure tightness. Practical actions are briefly indicated. Key points indicating the feasibility of a strategy of containment in vessel are discussed. The impact of reactor power on the robustness of an approach with containment in vessel is also discussed. An overview of technological evolutions and contributions of researches made by the IRSN is finally proposed

  17. 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

  18. 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

  19. 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.)

  20. 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.)

  1. 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

  2. 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

  3. Reactor Containment Spray Technology Program

    Energy Technology Data Exchange (ETDEWEB)

    Row, T. H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    1968-12-15

    The design basis accident in water moderated power reactors is a loss-of-coolant accident in which water sprays are generally employed to control the containment pressure transient by condensing the released steam-air mixture. Additives to the spray have been proposed as a way to increase their usefulness by enhancing the removal of various forms of radioiodine from the containment atmosphere. A program to investigate the gas-liquid systems involved is co-ordinated by ORNL for the US Atomic Energy Commission. A basic part of the program is the search for various chemical additives that will increase the spray affinity for molecular iodine and methyl iodide. A method for evaluating additives was developed that measures equilibrium distribution coefficients for iodine between air and aqueous solutions. Additives selected are used in single drop-wind tunnel experiments where the circulating gas contains iodine or CH{sub 3}I. Mass transfer coefficients and transient distribution coefficients have been determined as a function of relative humidity, temperature, drop size, and solution pH and concentration. Tests have shown that surfactants and organic amines increase the solution ability to getter CH{sub 3}l. Results from single drop tests help in planning spray experiments in the Nuclear Safety Pilot Plant, a large ({approx}38 m{sup 3}) facility, where accident conditions are closely simulated. Iodine and CH{sub 3}I removal rates have been determined for a number of solutions, including 1 wt% Na{sub 2}S{sub 2}O{sub 3} + 3000 ppm B + 0.153 M NaOH and 3000 ppm B + 0.153 M NaOH. The additive has very little effect in removal of I{sub 2} with half-lives of less than 1 mm typical for any aqueous solution. These same solutions remove CH{sub 3}I with a half-life of one hour. Analytical models for the removal processes have been developed. Consideration is also being given to corrosion, thermal and radiation stability of the solutions. Radiation studies have indicated the loss

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.)

  10. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi; Masumi, Ryoji; Soneda, Hideo.

    1994-01-01

    A fuel assembly comprises a plurality of fuel rods filled with nuclear fuels, a plurality of burnable poison rods incorporated with burnable poisons, and water rods which can vary the height in the tube depending on the coolant flow rate flown into the assembly. The amount of entire burnable poisons of the burnable poison-containing rods in adjacent with the water rods is smaller than the amount of entire burnable poisons in the burnable poison containing rods not in adjacent with the water rods. Then the average concentration of burnable poisons in the axial upper half region is made smaller than the average concentration of the burnable poisons at the axial lower half region. Further, a burnable poison concentration at the upper half region of at least one of burnable poison-containing rods in adjacent with the water rods is made lower than the burnable poison concentration in the lower half region. Since this can fasten the combustion of the burnable poisons, a fuel assembly having good fuel economy can be attained. (I.N.)

  11. Containment vessel construction for nuclear power reactors

    International Nuclear Information System (INIS)

    Sulzer, H.D.; Coletti, J.L.

    1975-01-01

    A nuclear containment vessel houses an inner reactor housing structure whose outer wall is closely spaced from the inner wall of the containment vessel. The inner reactor housing structure is divided by an intermediate floor providing an upper chamber for housing the reactor and associated steam generators and a lower chamber directly therebeneath containing a pressure suppression pool. Communication between the upper chamber and the pressure suppression pool is established by conduits extending through the intermediate floor which terminate beneath the level of the pressure suppression pool and by inlet openings in the reactor housing wall beneath the level of the pressure suppression pool which communicate with the annulus formed between the outer wall of the reactor housing structure and the inner wall of the containment vessel. (Official Gazette)

  12. 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.)

  13. 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.)

  14. Simplified safety and containment systems for the iris reactor

    International Nuclear Information System (INIS)

    Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

  15. Containment shells of reactor compartments at foreign NPPs

    International Nuclear Information System (INIS)

    Demidov, A.P.; Savchenko, V.A.

    1989-01-01

    The modern designes of containment shells (CS) of NPP reactor compartments is described. Much attention is paid to the PCS-3 project envisaging CS inclusion in the complex of NPP passive safety system. The PCS-3 system is developed in the USA for NPP with the improved PWR type reactor. The above system permits to cool the core quickly, to reduce steam pressure in CS down to a safe level and to prevent the discharge of radioactive products in the atmosphere in the case of accidents, even very serious, caused by loss of coolant and core dryout

  16. 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

  17. 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.)

  18. 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.)

  19. 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

  20. 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

  1. 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.

  2. Ventilation air conditioner for a reactor container

    International Nuclear Information System (INIS)

    Ikegame, Noboru; Nakagawa, Takeshi.

    1980-01-01

    Purpose: To suppress the variations in the internal pressure of a reactor container and smoothly ventilate the reactor container. Constitution: The air conditioner provides an air-flow-rate-control damper, a purge-air supply fan, and a filter device in the air-supply pipe of a reactor container. Furthermore, it provides a pressure difference detector at a part of the container. The air-flow-rate-control damper is connected electrically through a position-modulator-comparison amplifier to the pressure difference detector. When the filtration becomes insufficient by clogging of the filter device and the internal pressure increased abruptly in the container, the pressure-difference detector can detect it, and the damper is operated by a pressure regulator and the comparator so as to control the air flow to the container. Thus, the internal pressure variation is controlled so as to easily ventilate the container. (J.P.N.)

  3. 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

  4. 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

  5. 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

  6. A method of installing a reactor container

    International Nuclear Information System (INIS)

    Hayashi, Kenji; Murakawa, Hisao.

    1975-01-01

    Object: To achieve exact installation of a reactor container at a site. Structure: A pole is set upright at the center of a cylindrical base portion, a plurality of beams are disposed around the pole in a radial fashion to form a cone, a plurality of steel plates are mounted successively around the cone through a ring, and the steel plates are welded to each other to assemble and install a reactor container at the same time. (Kamimura, M.)

  7. 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

  8. 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.

  9. Actions to reduce radioactive emissions: prevention of containment failure by flooding Containment and Reactor Cavity

    International Nuclear Information System (INIS)

    Fornos Herrando, J.

    2013-01-01

    The reactor cavity of Asco and Vandellos II is dry type, thus a severe accident leading to vessel failure might potentially end up resulting in the loss of containment integrity, depending on the viability to cool the molten core. Therefore, significant radioactive emissions could be released to outside. In the framework of Fukushima Stress Tests, ANAV has analyzed the convenience of carrying out different actions to prevent failure of the containment integrity in order to reduce radioactive emissions. The aim of this paper is to present and describe the main phenomenological aspects associated with two of these actions: containment flooding and reactor cavity flooding.

  10. 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)

  11. Leak monitoring method for a reactor container

    International Nuclear Information System (INIS)

    Uehara, Toshio.

    1987-01-01

    Purpose: To confirm leakages from a container upon nuclear reactor operation. Method: Leakages from a nuclear reactor container has been prevented by lowering the inner pressure of the container relative to the external pressure. In the conventional method of calculating the leakage by applying an inner pressure to the container and measuring the pressure change, etc. after the elapse of a pre-determined time, the measurement has to be conducted at periodical inspection when the nuclear reactor is shut-down. In view of the above, the leak test is conducted in the present invention by applying a slight inner pressure to the inside of the reactor container by supplying gases from a gas supply system and detecting the flow rate of the gases in the gas supply system while maintaining the slight inner pressure constant by controlling the supply and discharge of the gases. By applying such a inner pressure as causing no effect to the reactor operation, it is possible to monitor the leaks during operation and to detect the flow rate value surely and continuously if the leak is resulted. (Kamimura, M.)

  12. 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.)

  13. Containment and surveillance techniques at power reactors

    International Nuclear Information System (INIS)

    Stirling, A.J.

    1982-01-01

    This session will provide participants with an understanding of the functions of safeguards equipment at power reactors, including equipment for fuel accounting, video and film surveillance, diversion monitoring, and containment and surveillance of irradiated fuel in storage. In addition, some appreciation of the impact that reactor safeguards have on the plant operator will be gained. From this, participants will be able to ensure that a reactor safeguards system meets their nation's international and national nonproliferation objectives with a minimum of interference to plant operations

  14. Fluid mixing in reactor containment

    International Nuclear Information System (INIS)

    Deoras M Prabhudharwadkar; Kannan N Iyer

    2005-01-01

    Full text of publication follows: Hydrogen release and distribution in nuclear power plant containment is an important safety issue. Selection of a proper turbulence model is important for accurate estimation of the mixing process. The selection of turbulence model is dictated by the best compromise between accuracy and computational efforts. For this, three different turbulence models, viz. Standard k-ε, RNG k-ε and Reynolds Stress Model, based on Reynolds averaged Navier Stokes equations (RANS) approach, were used. The computations were done using the CFD code FLUENT, which is based on the control volume methodology. The computational results were compared with the experimental results of HYMIS test facility, where helium was used to simulate hydrogen. The processes of helium plume rise, multiple plume merging, distribution and mixing were studied. Based on these computations, a simple analytical/empirical zone based model was formulated for the same problem, which predicted the helium concentration reasonably accurately and quickly. (authors)

  15. 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)

  16. 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.

  17. 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

  18. 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

  19. Containment for small pressurized water reactors

    International Nuclear Information System (INIS)

    Siler, W.C.; Marda, R.S.; Smith, W.R.

    1977-01-01

    Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

  20. 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

  1. 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...

  2. 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)

  3. 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

  4. Reactor container and controlling method thereof

    International Nuclear Information System (INIS)

    Hosaka, Seiichi.

    1990-01-01

    An object of the present invention is to prevent stress corrosion crack caused in pipelines made of stainless steels by preventing deposition of chlorine, etc. on the surface, etc. of the pipelines. That is, an internal evolving gas elimination system comprises a gas extraction device for extracting gases in the reactor container, an obstacle elimination device for eliminating obstacles contained in the extracted gases and an internal gas elimination device for eliminating internal evolving gases contained in the extracted gases. Further, gases in the upper portion of the reactor container are extracted and then the ingredients of the internal evolving gases contained in the gases are eliminated and, thereafter, the gases are supplied to the lower portion of the container to keep the relative humidity in the reactor container to less than 20%RH. As a result, since the internal evolving gases are eliminated and the relative humidity at the inside is kept to less than 20%RH, deposition of chlorine or salts on the pipelines can be prevented to thereby prevent the stress corrosion cracks. (I.S.)

  5. Containment concepts assessment for the SEAFP reactor

    International Nuclear Information System (INIS)

    Di Pace, L.; Natalizio, A.

    2000-01-01

    A simple methodology has been developed for making relative comparisons of potential containment designs for future fusion reactors. The assessment methodology requires only conceptual design information. The application of this methodology, at the early stages of a fusion reactor design, provides designers useful information regarding the suitability of various containment designs and design features. Because the radiation hazard from the operation of future fusion power reactors is expected to be low, the containment design, in addition to public safety, needs to take into account worker safety considerations, as well as factors important to the reliable and economical operation of the power plant. Several containment concepts have been assessed with a methodology that takes into account public safety, worker safety, operability and maintainability as well as cost. This paper describes this methodology and presents the results of the assessment. The paper concludes that, to obtain a containment design that is optimised with respect to safety, operational and cost factors, designers should focus on a containment that is conceptually simple-that is, one utilising a single, large containment building without relying on special features such as expansion volumes, pressure suppression pools or spray systems

  6. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  7. 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

  8. 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)

  9. 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

  10. Vent control device for nuclear reactor container

    International Nuclear Information System (INIS)

    Kubota, Ryuji.

    1989-01-01

    The present invention concerns automatic prevention of abnormal over-pressure and hydrogen gas flashing in a BWR type reactor container. That is, (1) if the pressure in the container is abnormally increased, the gas in the pressure suppression chamber is released to reduce the pressure thereby preventing over-pressure damage to the container. (2) Then, if exhaust gases are burnt to cause flashing explosion danger for the gases in the reactor container, the gas release is interrupted. The foregoing two functioins are automatically conducted in this device. Specifically, when the pressure in the reactor container reaches a predetermined allowable limit, a remote control operation valve is opened by automatic control means to release the gas in the vessel. Since the gas flow rate at the start of the release exceeds flame propagation velocity, there is no worry for flashing explosion. Further, if the pipeway flow velocity near the atmospheric release is reduced to less than the flame propagation velocity of the hydrogen gas, the opened valve is automatically closed. Accordingly, propagation of hydrogen gas flame into the container thus causing explosion can surely be prevented. (K.M.)

  11. Trends in the development of reactor containments

    International Nuclear Information System (INIS)

    Turricchia, A.

    1977-01-01

    The paper presents a review of the technical developments which the containment systems of BWRs, PWRs and HWRs have undergone in the USA, Europe and Canada. The great variety of existing containment systems have been classified for each type of reactor, according to: principle of operation (i.e. dry containment; pressure suppression containment; containment with vacuum building; sub-atmospheric containment; etc.); type and number of the physical barriers provided against the dispersion of fission products to the environment (single containment, double containment, single containment with local collection of the leaks from the most probable leakage paths, or with pressurization of the potential leakage paths, etc.); and structural characteristics of the containment barriers (reinforced concrete, prestressed concrete, metal containment, liners). In the framework of this classification the main features of the various containment systems are illustrated, with special emphasis on the most modern designs. The difference in trend among various countries and various designers are highlighted. A review is also made of the various containment auxiliary systems which would operate in accident conditions (isolation systems, heat removal system, atmosphere control system, etc.). Also for these systems the main design criteria are set forth and present trends are described. Finally, a brief survey is also made of the present trends with regards to the design of containment structures against external events be they of natural origin (earthquake, tornadoes, etc.) or connected with human activities (airplane crash, explosions, etc.) [fr

  12. 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

  13. 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

  14. Detonability of containment building atmospheres during core-meltdown accidents

    International Nuclear Information System (INIS)

    Jaung, R.; Berlad, L.; Pratt, W.

    1983-01-01

    During Core-Meltdown Accidents in Light Water Reactors, significant quantities of combustible gases could be released to the containment building. The highest possible peak pressure fields that may occur through combustion processes are associated with detonation phenomena. Accordingly, it is necessary to understand and identify the possible ways in which detonations may or may not occur. Although no comprehensive theory of detonation is currently available, there are useful guidelines, which can be derived from current theoretical concepts and the body of experimental data. This paper examines these guidelines and indicates how they may be used to evaluate the possible occurrence of detonation-related combustion processes. In particular, this study identifies three features that an initiation source must achieve if it is to ultimately result in a stable detonation. One of these features requires post-shock initial conditions that lead to very short ignition delays. This concept is used to examine the possibility of achieving quasi-steady detonation phenomena in nuclear reactor containment buildings during postulated core-melt accidents

  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. 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.

  18. The WINCON programme - validation of fast reactor primary containment codes

    International Nuclear Information System (INIS)

    Sidoli, J.E.A.; Kendall, K.C.

    1988-01-01

    In the United Kingdom safety studies for the Commercial Demonstration Fast Reactor (CDFR) include an assessment of the capability of the primary containment in providing an adequate containment for defence against the hazards resulting from a hypothetical Whole Core Accident (WCA). The assessment is based on calculational estimates using computer codes supported by measured evidence from small-scale experiments. The hydrodynamic containment code SEURBNUK-EURDYN is capable of representing a prescribed energy release, the sodium coolant and cover gas, and the main containment and safety related internal structures. Containment loadings estimated using SEURBNUK-EURDYN are used in the structural dynamic code EURDYN-03 for the prediction of the containment response. The experiments serve two purposes, they demonstrate the response of the CDFR containment to accident loadings and provide data for the validation of the codes. This paper summarises the recently completed WINfrith CONtainment (WINCON) experiments that studied the response of specific features of current CDFR design options to WCA loadings. The codes have been applied to some of the experiments and a satisfactory prediction of the global response of the model containment is obtained. This provides confidence in the use of the codes in reactor assessments. (author)

  19. 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

    accidents and, secondly, the physical phenomena, studies and analyses described in Chapters 5 to 8. Chapter 5 is devoted to describing the physical phenomena liable to occur during a core melt accident, in the reactor vessel and the reactor containment. It also presents the sequence of events and the methods for mitigating their impact. For each of the subjects covered, a summary of the physical phenomena involved is followed by a description of the past, present and planned experiments designed to study these phenomena, along with their modelling, the validation of which is based on the test results. The chapter then describes the computer codes that couple all of the models and provide the best current state of knowledge of the phenomena. Lastly, this knowledge is reviewed while taking into account the gaps and uncertainties, and the outlook for the future is presented, notably regarding experimental programmes and the development of modelling and numerical simulation tools. Chapter 6 focuses on the behaviour of the containment enclosures during a core melt accident. After summarising the potential leakage paths of radioactive substances through the different containments in the case of the accidents chosen in the design phase, it presents the studies of the mechanical behaviour of the different containments under the loadings that can result from the hazards linked with the phenomena described in Chapter 5. Chapter 6 also discusses the risks of containment building bypass in a core melt accident situation. Chapter 7 presents the lessons learned regarding the phenomenology of core melt accidents and the improvement of nuclear reactor safety. Lastly, Chapter 8 presents a review of development and validation efforts regarding the main computer codes dealing with 'severe accidents', which draw on and build upon the knowledge mainly acquired through the research programmes: ASTEC (IRSN and GRS), MAAP-4 (FAI (US)) and used by EDF and by utilities in many other

  20. 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.)

  1. Modeling of hydrogen stratification in a pressurized water reactor containment with the contain computer code

    International Nuclear Information System (INIS)

    Kljenak, I.; Skerlavaj, A.; Parzer, I.

    1999-01-01

    Hydrogen distribution during a severe accident in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN computer code. The accidents is initiated by a large-break loss-of-coolant accident which is nit successfully mitigated by the action of the emergency core cooling system. Cases with and without successful actuation of spray systems and fan coolers were considered. The simulations predicted hydrogen stratification within the containment main compartment with intensive hydrogen mixing in the containment dome region. Pressure and temperature responses were analyzed as well.(author)

  2. Models of iodine behavior in reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F.; Beahm, E.C.; Kress, T.S.

    1992-10-01

    Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents.

  3. Containment vessel for a nuclear reactor

    International Nuclear Information System (INIS)

    Yamanari, Sh.; Horiuchi, T.; Sugisaki, T.; Tominaga, K.

    1985-01-01

    A containment vessel for a nuclear reactor having a dry well for mounting therein a pressure vessel for containing the nuclear reactor, a pressure suppressing chamber having a pool of coolant therein, and a vent pipe device for releasing therethrough into the pool of coolant within the pressure suppressing chamber steam which will be produced as a result of the occurrence of an accident and escape into the dry well. The vent pipe device includes a plurality of vent pipe members inserted in the pool of coolant within the pressure suppressing chamber and each having at least one exhaust port opening in the coolant. The vent pipe members are divided into a plurality of groups in such a manner that the vent pipe members of different groups differ from one another in the length of submerged portions of the vent pipe members interposed between the liquid of the coolant within the pressure suppressing chamber and the exhaust ports of the vent pipe members

  4. Models of iodine behavior in reactor containments

    International Nuclear Information System (INIS)

    Weber, C.F.; Beahm, E.C.; Kress, T.S.

    1992-10-01

    Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents

  5. 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

  6. 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

  7. 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

  8. 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

  9. Venting device for nuclear reactor container

    International Nuclear Information System (INIS)

    Yamashita, Masahiro; Ogata, Ken-ichi.

    1994-01-01

    An airtight vessel of a venting device of a nuclear reactor container is connected with a reactor container by way of a communication pipeline. A feed water tank is disposed at a position higher than the liquid surface of scrubbing water in the airtight vessel for supplying scrubbing water to the airtight vessel. In addition, a scrubbing water storage tank is disposed at a position hither than the feed water tank for supplying scrubbing water to the feed water tank. Storage water in the feed water tank is introduced into the airtight vessel by the predetermined opening operation of a valve by the pressure exerted on the liquid surface and the own weight of the storage water. Further, the storage water in the scrubbing water storage tank is led into the feed water tank by the water head pressure. The scrubbing water for keeping the performance of the venting device of the reactor container can be supplied by a highly reliable method without using AC power source or the like as a driving source. (I.N.)

  10. 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

  11. Pipe line construction for reactor containment buildings

    International Nuclear Information System (INIS)

    Aoki, Masataka; Yoshinaga, Toshiaki

    1978-01-01

    Purpose: To prevent the missile phenomenon caused by broken fragments due to pipe whip phenomenon in a portion of pipe lines connected to a reactor containment from prevailing to other portions. Constitution: Various pipe lines connected to the pressure vessel are disposed at the outside of the containments and they are surrounded with a plurality of protection partition walls respectively independent from each other. This can eliminate the effect of missile phenomena upon pipe rupture from prevailing to the pipe lines and instruments. Furthermore this can afford sufficient spaces for the pipe lines, as well as for earthquake-proof supports. (Horiuchi, T.)

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.)

  17. 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)

  18. 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.)

  19. 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

  20. 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

  1. 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

  2. Pressure suppression facility for reactor container

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  3. Studies for reactor containment at Narora

    International Nuclear Information System (INIS)

    Kurien, M.Z.

    1975-01-01

    The primary reactor containment at Narora has the unusual function of also carrying the heavy boilers (heat exchangers) and providing a flat work surface over the roof. The site being in a seismically active region, certain conflicting requirements in regard to ideal conditions to resist and analyse functional loads and seismic loads have to be reconciled. Alternative schemes are proposed and the stress flow for each major loading case examined and their consequent relative advantages discussed. Problems at the joints and functions are examined and the implications of various arrangements are reviewed. Suggestions to reduce the magnitude of the problems are tentatively proposed. (author)

  4. 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

  5. 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

  6. Tritium containment of controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

    1979-01-01

    It is well known that tritium is used as the fuel for nuclear fusion reactors. The neutrons produced by the nuclear fusion reaction of deuterium and tritium react with lithium in blankets, and tritium is produced. The blankets reproduce the tritium consumed in the D-T reaction. Tritium circulates through the main cooling system and the fuel supply and evacuation system, and is accumulated. Tritium is a radioactive substance emitting β-ray with 12.6 year half-life, and harmful to human bodies. It is an isotope of hydrogen, and apt to diffuse and leak. Especially at high temperature, it permeates through materials, therefore it is important to evaluate the release of tritium into environment, to treat leaked tritium to reduce its release, and to select the method of containing tritium. The permeability of tritium and its solubility in structural materials are discussed. The typical blanket-cooling systems of nuclear fusion reactors are shown, and the tungsten coating of steam generator tubes and tritium recovery system are adopted for reducing tritium leak. In case of the Tokamak type reactor of JAERI, the tritium recovery system is installed, in which the tritium gas produced in blankets is converted to tritium steam with a Pd-Pt catalytic oxidation tower, and it is dehydrated and eliminated with a molecular sieve tower, then purified and recovered. (Kako, I.)

  7. 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)

  8. Nuclear reactor melt-retention structure to mitigate direct containment heating

    Science.gov (United States)

    Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

    1991-01-01

    A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

  9. 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

  10. 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.

  11. 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)

  12. 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)

  13. 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)

  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. 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.

  16. 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

  17. 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.)

  18. 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.)

  19. 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).

  20. 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.

  1. 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.)

  2. 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.

  3. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  4. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  5. 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...

  6. Heat removing device for nuclear reactor container facility

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-09-30

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

  7. 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.

  8. Optically monitoring device in reactor container

    International Nuclear Information System (INIS)

    Takeuchi, Tsutomu; Kawamoto, Kikuo.

    1993-01-01

    In the device of the present invention, cable penetrations are necessary for transmission from a great number of electrical instrumentations disposed in a reactor container to the outside. Optical cables are passed through the cable penetrations to use optical signals for signal transmission. That is, pulses are injected from one end of the optical cable and a specific light (raman scattered light) among reflected lights, after elapse of a predetermined time, is measured. With such procedures, temperature at the reflection point can be measured. In this case, if emission and discrimination of the pulses are conducted in a time sharing fashion, the temperature is measured as an average value for the 1m length corresponding to the time determined by the limit. Accordingly, greater number of temperature measuring points than that in the prior art (there is a reactor which has about 170 points) are measured by a lesser number of cables (one at minimum). For instruments used for other than temperature, if the device is diagnosed by itself, by making the constitution intelligent and utilizing optical output, reliability for the measurement can be improved. (I.S.)

  9. 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.

  10. Hydrogen management in nuclear reactor containment

    International Nuclear Information System (INIS)

    Iyer, Kannan

    2014-01-01

    The talk will present the systematic methodology evolved to assess the hydrogen management in nuclear reactor containment during a severe accident. The focus is on the methodology evolved as the full problem is yet to be solved completely. First, the method to quantify mixing of hydrogen is presented. It is demonstrated that buoyancy modified model is adequate to quantify the process satisfactorily. On noting that the hydrogen levels are higher than the safe limits, effort was directed towards mitigating the concentration. Passive Auto-catalytic Recombiners (PAR) were identified as the potential devices for mitigation. Efforts were then directed to model these and a satisfactory one-step reaction derived from a 12 reaction model was evolved. This model was satisfactory when compared with experimental results with hydrogen concentration below 4%. However, the same when extended to hydrogen concentration of 20%, predicts very high concentration thereby indicating the need for experiments at high hydrogen concentration. (author)

  11. Filtered atmospheric venting of light water reactor containments

    International Nuclear Information System (INIS)

    Hedgran, A.; Ahlstroem, P.E.; Nilsson, L.; Persson, Aa.

    1982-11-01

    The aim of filtered venting is to improve the function of the reactor containment in connection with very severe accidents. By equipping the containment with a safety valve for pressure relief and allowing the released gases to pass through an effective filter, it should be possible to achieve a considerable protective effect. The work has involved detailed studies of the core meltdown sequence, how the molten core material runs out of the reactor vessel, what effect it has on concrete and other structures and how final cooling of the molten core material takes place. On the basis of previous Swedish studies, the project has chosen to study a filter concept that consists of a gravel bed of large volume. This filter plant shall not only retain the radioactive particles that escape from the containment through the vent line, but shall also condense the accompanying steam. After the government decided in 1981 that Barsebaeck was to be equipped with filtered venting and issued specifications regarding its performance, the project aimed at obtaining results that could be used to design and verify a plant for filtered venting at the Barsebaeck nuclear power station. As far as the other Swedish nuclear power plants at Oskarshamn, Ringhals and Forsmark are concerned, the results are only applicable to a limited extent. Additional studies are required for these nuclear power plants before the value of filtered venting can be assessed. Based on the results of experiments and analyses, the project has made a safety analysis with Barsebaeck as a reference plant in order to study how the introduction of filtered venting affects the safety level at a station. In summary, the venting function appears to entail a not insignificant reduction of risks for boiling water reactors of the Barsebaeck type. For a number of types of such very severe core accident cases, the filter design studied ensures a substantial reduction of the releases. However it has not been possible within the

  12. 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)

  13. Transients in a reactor containment after a HCDA

    International Nuclear Information System (INIS)

    Ghosh, A.K.

    1984-01-01

    The consequences of a hypothetical core meltdown accident in a fast reactor is analysed. Shock waves are generated in the surrounding medium after the energy release, which is assumed to be instantaneous and at a point. After discussing the difference in the predicted and experimentally observed peak pressure a semi-empirical approach is taken to arrive at a better estimate. This defines the loading on the containment, which is idealised as a combination of a shell and a plate. To simplify the analysis the shell is assumed to be in plane strain and axial symmetry is assumed for both components. The shell response is evaluated by using elasticity theory. Numerical results are presented for peak overpressure and pressure-time history and dynamic stresses in the containment. (Author) [pt

  14. 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

  15. 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

  16. 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)

  17. 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)

  18. Permanent cavity seal ring for a nuclear reactor containment arrangement

    International Nuclear Information System (INIS)

    Swidwa, K.J.; Salton, R.B.; Marshall, J.R.

    1990-01-01

    This patent describes a nuclear reactor containment arrangement. It comprises: a reactor pressure vessel which thermally expands and contracts during cyclic operation of the reactor, the vessel having a peripheral wall and a horizontally outwardly extending flange thereon; a containment wall having a shelf, the wall spaced from and surrounding the peripheral wall of the reactor pressure vessel defining an annular expansion gap therebetween, and an annular ring seal extending across the annular expansion gap to provide a water-tight seal therebetween

  19. 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

  20. 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

  1. 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)

  2. 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.)

  3. 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)

  4. 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

  5. Nuclear reactor melt-retention structure to mitigate direct containment heating

    International Nuclear Information System (INIS)

    Tutu, N.K.; Ginsberg, T.; Klages, J.R.

    1991-01-01

    This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  12. 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

  13. Course of pin fuel test In WWR-M reactor core

    International Nuclear Information System (INIS)

    Zakharov, A.S.; Kirsanov, G.A.; Konoplev, K.A.

    2005-01-01

    Pin type fuel element (FE) of square form with twisted ribs was developed in VNIINM as an alternative for tube type FE of research reactors. Two variants of full-scale fuel assemblies (FA) are under test in the core of PNPI WWR-M reactor. One FA contains FE with UO 2 LEU and other - UMo LEU. Both types of FE have an aluminum matrix. Results of the first stages of the test are presented. (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. 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

  16. 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

  17. 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)

  18. 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)

  19. 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.)

  20. 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)

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. Bounding analysis of containment of high pressure melt ejection in advanced light water reactors

    International Nuclear Information System (INIS)

    Additon, S.L.; Fontana, M.H.; Carter, J.C.

    1990-01-01

    This paper reports on the loadings on containment due to direct containment heating (DCH) as a result of high pressure melt ejection (HPME) in advanced light water reactors (ALWR) which were estimated using conservative, bounding analyses. The purpose of the analyses was to scope the magnitude of the possible loadings and to indicate the performance needed from potential mitigation methods, such as a cavity configuration that limits energy transfer to the upper containment volume. Analyses were performed for three cases which examined the effect of availability of high pressure reactor coolant system water at the time of reactor vessel melt through and the effect of preflooding of the reactor cavity. The amount of core ejected from the vessel was varied from 100% to 0% for all cases. Results indicate that all amounts of core debris dispersal could be accommodated by the containment for the case where the reactor cavity was preflooded. For the worst case, all the energy from in-vessel hydrogen generation and combustion plus that from 45% of the entire molten core would be required to equilibrate with the containment upper volume in order to reach containment failure pressure

  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. Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

    1991-01-01

    The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

  8. 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

  9. Degraded core accidents: review of aerosol behaviour in the containment of a PWR

    International Nuclear Information System (INIS)

    Nichols, A.L.; Walker, B.C.

    1981-09-01

    Low probability-high consequence accidents have become an important issue in reactor safety studies. Such accidents would involve damage to the core and the subsequent release of radioactive fission products into the environment. Aerosols play a major role in the transport and removal of these fission products in the reactor building containment. The aerosol mechanisms, computer modelling codes and experimental studies used to predict aerosol behaviour in the containment of a PWR are reviewed. There are significant uncertainties in the aerosol source terms and specific recommendations have been made for further studies, particularly with respect to code development and high density aerosol-fission product transport within closed systems. (author)

  10. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  11. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

  12. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

    International Nuclear Information System (INIS)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects

  13. Safety assessment to support NUE fuel full core implementation in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fan, H.Z.; Laurie, T.; Siddiqi, A.; Li, Z.P.; Rouben, D.; Zhu, W.; Lau, V.; Cottrell, C.M. [CANDU Energy Inc., Mississauga, Ontario (Canada)

    2013-07-01

    The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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)

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

    International Nuclear Information System (INIS)

    West, C.D.

    1995-01-01

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

  1. 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).

  2. 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)

  3. 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)

  4. 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

  5. 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.)

  6. 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

  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. 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)

  9. 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).

  10. 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

  11. Evaluation of final vapor pressures in the loss of flow accident in an irradiation device of a pool reactor core

    International Nuclear Information System (INIS)

    Verri, A.

    1987-01-01

    The reliability feature, are described for a device containing samples, at a temperatures of 300 grade centigrades, in a reactor core for a long time. After an examination of the maximum accident event, the maximum vapour pressure originated by the inlet of reactor cooling water into the experimental device, is evaluated

  12. 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

  13. 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.

  14. 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

  15. 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.)

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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.

  4. 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

  5. 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.)

  6. Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

    International Nuclear Information System (INIS)

    1992-04-01

    Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

  7. 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.)

  8. 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

  9. 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)

  10. Shielded transport containers for reactor waste

    International Nuclear Information System (INIS)

    Grundfelt, B.; Eriksson, E.

    The report presents that part of risk analysis which deals with the frequency of breakdowns and the damage on containers. The report focusses on shielded containers made of reinforced concrete. Also a container made of steel is referred to the cases of breakdown are closely allied to collisions with ships. The frequency of breakdowns which might damage the containers is low in all respects, namely 1.10 -5 per year or lower for the shielded container. (G.B.)

  11. 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.

  12. 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)

  13. 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)

  14. An evaluation of designed passive Core Makeup Tank (CMT) for China pressurized reactor (CPR1000)

    International Nuclear Information System (INIS)

    Wang, Mingjun; Tian, Wenxi; Qiu, Suizheng; Su, Guanghui; Zhang, Yapei

    2013-01-01

    Highlights: ► Only PRHRS is not sufficient to maintain reactor safety in case of SGTR accident. ► The Core Makeup Tank (CMT) is designed for CPR1000. ► Joint operation of PRHRS and CMT can keep reactor safety during the SGTR transient. ► CMT is a vital supplement for CPR1000 passive safety system design. - Abstract: Emergency Passive Safety System (EPSS) is an innovative design to improve reliability of nuclear power plants. In this work, the EPSS consists of secondary passive residual heat removal system (PRHRS) and the reactor Core Makeup Tank (CMT) system. The PRHRS, which has been studied in our previous paper, can effectively remove the core residual heat and passively improve the inherent safety by passive methods. The designed CMT, representing the safety improvement for CPR1000, is used to inject cool boron-containing water into the primary system during the loss of coolant accident. In this study, the behaviors of EPSS and transient characteristics of the primary loop system during the Steam Generator Tube Rupture (SGTR) accident are investigated using the nuclear reactor thermal hydraulic code RELAP5/MOD3.4. The results show that the designed CMT can protect the reactor primary loop from boiling and maintain primary loop coolant in single phase state. Both PRHRS and CMT operation ensures reactor safety during the SGTR accident. Results reported in this paper show that the designed CMT is a further safety improvement for CPR1000

  15. Thermohydraulic design of the CAREM reactor's containment

    International Nuclear Information System (INIS)

    Abbate, P.

    1990-01-01

    The analysis made for the calculation of the temperature and pressure basic charges affecting the CAREM 25 reactor's contention system is presented. The case of a primary cooler loss, with simultaneous loss of electric supply, is analyzed. Different aspects of the numerical model used, are discussed. (Author) [es

  16. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  17. 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

  18. 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.)

  19. Excess-pressure suppression device in a reactor container

    International Nuclear Information System (INIS)

    Nishio, Masahide

    1985-01-01

    Purpose: To reliably decrease the radioactivity of radioactive gases when they are released externally. Constitution: The exit of a gas exhaust pipe for discharging gases in a reactor container, on generation of an excess pressure in the reactor container upon loss of coolant accident, is adapted to be always fluided in the cooling tank. Then, the exhaust gases discharged in the cooling tank is realeased to the atmosphere. In this way, the excess pressure in the reactor container can be prevented previously and the radioactivity of the gases released externally is significantly reduced by the scrubbing effect. (Kamimura, M.)

  20. Method of judging leak sources in a reactor container

    International Nuclear Information System (INIS)

    Maeda, Katsuji.

    1984-01-01

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

  1. Materials considerations for UF6 gas-core reactor. Interim report for preliminary design study

    International Nuclear Information System (INIS)

    Wagner, P.

    1977-04-01

    The limiting materials problem in a high-temperature UF 6 core reactor is the corrosion of the core containment vessel. The UF 6 , the lower fluorides of uranium, and the fluorine that exist at the anticipated reactor operating conditions (1000 K and about one atmosphere UF 6 ) are all corrosive. Because of this, the materials evaluation effort for this reactor design study has concentrated on the identification of a viable system for the containment vessel that meets both the materials and neutronic requirements. A study of the literature has revealed that the most promising corrosion-resistant candidates are Ni or Ni-Al alloys. One of the conclusions of this work is that the containment vessel use a nickel liner or clad since the use of Ni as a structural member is precluded by its relative blackness to thermal neutrons. Estimates of corrosion rates of Ni and Ni-Al alloys, the effects of the pressure and temperature of F 2 on the corrosion rates, calculated equilibrium gas compositions at reactor core operating conditions, suggested methods of fabrication, and recommendations for future research and development are included

  2. 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.)

  3. 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)

  4. 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

  5. Experimental study on air ingress during a primary pipe rupture accident with a graphite reactor core simulator

    International Nuclear Information System (INIS)

    Takeda, Tetsuaki; Hishida, Makoto; Baba, Shinichi

    1991-11-01

    When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary cooling system reduces the pressure. After the pressures are balanced between the reactor and the container, air is expected to enter into the reactor core from the breach. It seems to be probable that the graphite structures is oxidized by air. Hence, it is necessary to investigate the air ingress process and the behavior of the generating gases by the oxidation reactions. The previous experimental study is performed on the molecular diffusion and natural convection of the two component gas mixtures using a test model simulating simply the reactor. Objective of the study was to investigate the air ingress process during the early stage of the primary pipe rupture accident. However, since the model did not have any kind of graphite components, the reaction between graphite and oxygen was not simulated. The present model includes the reactor core and the high temperature plenum simulators made of graphite. The major results obtained in the present study are summarized in the followings: (1) The air ingress process with graphite oxidation reaction is similar to that without the reaction qualitatively. (2) When the reactor core simulator is maintained at low temperatures (lower than 450degC), the initiation time of the natural circulation of air is almost equal to that of the natural circulation of nitrogen. On the other hand, when the temperature of the reactor core simulator is high (more than 500degC), the initiation time of the natural circulation of air is earlier than that of nitrogen. (3) When the temperature of the reactor core simulator is higher than 600degC, oxygen is almost dissipated by the graphite structures. When the temperature of the reactor core simulator is below 700degC, carbon dioxide mainly is generated by the oxidation reactions. (author)

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. Measuring technique of super high temperature thermal properties of reactor core materials

    International Nuclear Information System (INIS)

    Ono, Akira; Baba, Tetsuya; Watanabe, Hideo; Matsumoto, Tsuyoshi

    1998-01-01

    In this study, thermal properties of reactor core materials used for water cooled reactors and FBR were tried to develop a technique to measure their melt states at less than 3,000degC in order to contribute more correct evaluation of the reactor core behavior at severe accident. Then, a thermal property measuring method of high temperature melt by using floating method was investigated and its fundamental design was begun to investigate under a base of optimum judgement on the air flow floating throw-down method. And, in order to measure emissivity of melt specimen surface essential for correct temperature measurement using the throw down method, a spectroscopic emissivity measuring unit using an ellipsometer was prepared and induced. On the thermal properties measurement using the holding method, a specimen container to measure thermal diffusiveness of the high temperature melts by using laser flashing method was tried to prepare. (G.K.)

  11. 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

  12. 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

  13. Passive containment system for a nuclear reactor

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1976-01-01

    A containment system is described that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is continuously maintained submerged in liquid. The primary containment vessel is restored to a high subatmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means

  14. 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

  15. Corrosion of cermet cores of fuel plates for nuclear research reactor

    International Nuclear Information System (INIS)

    Durazzo, M.; Ramanathan, L.V.

    1984-01-01

    Materials Testing Reactor (MTR) type fuel plates containing U 3 O 8 -Al cores and clad with Al are used in various research reactor. Preliminary investigations, where in the cladding of samples was drilled to simulate conditions of rupture due to pitting attack, revealed that considerable quantities of H 2 was evolved upon exposure of the core to water. The corrosion of cermets cores of different densities was characterized as a function of H 2 evolution that revealed 3 stages. A first stage consisting of an incubation period followed by initiation of H 2 evolution, a second stage with a constant rate of H 2 evolution and a third stage with a low rate of H 2 evolution. All 3 stages were found to vary as a function of cermet density and water temperature. (Author) [pt

  16. 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

  17. Neutronic and mechanical design of the reactor core of the Opus system

    Energy Technology Data Exchange (ETDEWEB)

    Raepsaet, X.; Pascal, S. [CEA Saclay, Dept. Modelisation de Systemes et Structures (DEN/DM2S), 91 - Gif sur Yvette (France)

    2007-07-01

    Since a few years now, Cea decided to maintain a waking state in its space nuclear activities by carrying out some conceptual studies of embarked nuclear power systems in the range of 100-500 kWe. Results stemming from these ongoing studies are gathered in the project OPUS -Optimized Propulsion Unit System-. This nuclear power system relies on a fast gas-cooled reactor concept coupled either to a Brayton cycle or to a more ambitious energy conversion system using a Hirn cycle to dramatically reduce the size of the radiator. The OPUS reactor core consists of an arrangement of enriched graphite elements of hexagonal cross-section. Their length is equal to the core diameter (48 cm). Coated fuel particles containing enriched (93%) uranium are embedded in these fuel elements. Each fuel element is designed with a centered axial channel through which flows the working fluid: a mixture of helium and xenon gas. This reactor is expected to have an operating life of over 2000 days at full power. In fact the main questions remain on the fuel element manufacturing and on the mechanical design (type and size of particles, packing fraction in the matrix, final core diameter and mass). Especially, the nuclear reactor has been defined considering the possible synergies with the next generation of terrestrial nuclear reactor (International Generation IV Forum). Based on relatively short-term technologies, the same reactor is designed to cover a wide range of power: 100 to 500 kWe without core design modification. The final reactor design presented in this paper is the result of a coupled analysis between the thermomechanical and the neutronic aspects.

  18. 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

  19. 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)

  20. 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)

  1. 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

  2. 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

  3. 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.)

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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

  9. Collecting and recirculating condensate in a nuclear reactor containment

    International Nuclear Information System (INIS)

    Schultz, T.L.

    1993-01-01

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures

  10. Collecting and recirculating condensate in a nuclear reactor containment

    Science.gov (United States)

    Schultz, Terry L.

    1993-01-01

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

  11. 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.)

  12. Description of the containment for a stationary pressurized water reactor

    International Nuclear Information System (INIS)

    Hermani, S.

    1986-01-01

    The function of the containment is to prevent the inadvertent release of radioactive fission products from the reactor coolant system to the atmosphere and to provide biological shielding during both normal and accident operation. Basically three different containment concepts 1) the dry containment, 2) the subatmospheric containment, and 3) the ice condenser containment, have been developed, based on how the accident energy release from the reactor coolant system is controlled. The containment structure can be either 1) reinforced concrets with inside liner, 2) prestressed concrete with inside, or 3) full steel cylinder or steel sphere with separate concrete shield. The size of the containment is largely dictated by the required net free volume, that satisfies the energy release criteria due to the design basic accident. The design and construction methods applied to this structure guarantee that the containment will carry out its safety function. This was proven by the Three Mile Island accident. (author)

  13. 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.)

  14. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-07-01

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

  16. 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)

  17. Pressurized Water Reactor containment in Russia

    International Nuclear Information System (INIS)

    Taymouri, Majid.

    1993-01-01

    One of the most important systems of nuclear power plants from an economical point of view and view point of safety is containment; Therefore, the containments designed in Russia were studied in the first chapter. Russian general rules and requirements of structure of accident localization system were illustrated. Methods of accident localization system rooms tested for tightness and strength are presented in chapter three. Russian specialists have been working hard to ensure the safety culture in building structures and operational procedures and the have successfully implemented these objectives in new nuclear power plant designs and rules

  18. 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)

  19. 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)

  20. 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

  1. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

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

  3. 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

  4. 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

  5. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  6. 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

  7. 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.)

  8. 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

  9. Leakage detection system in nuclear reactor container

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The present invention comprises an injection means for adding radioactive materials to coolants in a container cooler, a gamma ray amplitude analyzer connected to coolant pipelines and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to a drain water sump and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to various kinds of pipelines and a means for recording/transferring the data of the result of the measurement, and a data processing means for comparing and analyzing the measured data of each of the gamma ray amplitude analyzers inputted from each of date recording/transferring means. The gamma ray amplitude analysis for each of the pipelines and drain water sump are conducted at an appropriate frequency, and the measured data are compared and analyzed, to improve the detection accuracy for a trace amount of leakage from each of the pressure pipelines and the container cooler coolant pipelines, thereby enabling to specify the pipeline having leakage. Maintenance efficiency is improved, and severe rupture accident in each of pressure pipelines is prevented previously. (N.H.)

  10. Liquid metal reactor applications of the CONTAIN code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Gido, R.; Valdez, G.D.; Scholtyssek, W.

    1988-01-01

    The CONTAIN code is the NRC's best-estimate code for the evaluation of the conditions that may exist inside a reactor containment building during a severe accident. Included in the phenomena modeled are thermal-hydraulics, radiant and convective heat transfer, aerosol loading and transient response, fission product transport and heating effects, and interactions of sodium and corium with the containment atmosphere and structures. CONTAIN has been used by groups in Japan and West Germany to assess its ability to analyze accident consequences for liquid metal reactor (LMR) plants. In conjunction with this use, collaborative efforts to improve the modeling have been pursued. This paper summarizes the current state of the version of CONTAIN that has been enhanced with extra capabilities for LMR applications. A description of physical models is presented, followed by a review of validation exercises performed with CONTAIN. Some demonstration calculations of an integrated LMR application are presented

  11. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

    1995-01-01

    The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

  12. Lighting system for the lower core plate of a nuclear reactor

    International Nuclear Information System (INIS)

    Feuillet, P.; Bonin, J.P.

    1986-01-01

    The invention proposes a grazing lighting system for the lower core plate, creating an excellent contrast and offering a good estimation of the relief; it can stay at the same place during the whole or at least the greater part of the core refueling operation. This lighting system is proposed for a reactor of which the lower core plate has fuel assembly centering elements. It has a sealed vessel with a transparent side wall containing several lights independently controlled and each one illuminating a sector of its wall. The vessel has a bottom aimed at resting on the lower plate and provided with centering and holding means acting with several of the said centering means through the plate, and/or apertures for coolant through the plate, and an upper container provided with gripping and handling elements and sealed conduits for electrical cables feeding the lights [fr

  13. 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

  14. A 1055 ft/sec impact test of a two foot diameter model nuclear reactor containment system without fracture

    Science.gov (United States)

    Puthoff, R. L.

    1972-01-01

    A study to determine the feasibility of containing the fission products of a mobile reactor in the event of an impact is presented. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block at 1055 ft/sec. The model was significantly deformed and the concrete block demolished. No leaks were detected nor were any cracks observed in the model after impact.

  15. Performance tests of the reactor containment structures of HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kawaji, Satoshi; Iyoku, Tatsuo

    1998-03-01

    The containment structures of the HTTR consist of the reactor containment vessel (CV), service area (SA) and emergency air purification system, which minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. The CV is designed to withstand the temperature and pressure transients and to be leak-tight within the specified leakage limit even in the case of a rupture of the primary concentric hot gas duct. The pressure of inside of the SA should be maintained slightly lower than that of atmosphere by the emergency air purification system. The radioactive materials are released from the stack to environment via the emergency air purification system under the accident condition. Then the emergency air purification system should remove airborne radio-activities and should maintain proper pressure in the SA. We established the method to measure leak rate of the CV with closed reactor coolant pressure boundary although it is normally measured under opened reactor coolant pressure boundary as employed in LWRs. The CV leak rate test was carried out by the newly developed method and the expected performance was obtained. The SA and emergency air purification system were also confirmed by the performance test. We concluded that the reactor containment structures were fabricated to minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. (author)

  16. 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)

  17. 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

  18. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2017-05-15

    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  19. 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)

  20. 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

  1. 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.)

  2. Integrated analysis of core debris interactions and their effects on containment integrity using the CONTAIN computer code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Williams, D.C.; Tills, J.L.; Valdez, G.D.

    1987-01-01

    The CONTAIN computer code includes a versatile system of phenomenological models for analyzing the physical, chemical and radiological conditions inside the containment building during severe reactor accidents. Important contributors to these conditions are the interactions which may occur between released corium and cavity concrete. The phenomena associated with interactions between ejected corium debris and the containment atmosphere (Direct Containment Heating or DCH) also pose a potential threat to containment integrity. In this paper, we describe recent enhancements of the CONTAIN code which allow an integrated analysis of these effects in the presence of other mitigating or aggravating physical processes. In particular, the recent inclusion of the CORCON and VANESA models is described and a calculation example presented. With this capability CONTAIN can model core-concrete interactions occurring simultaneously in multiple compartments and can couple the aerosols thereby generated to the mechanistic description of all atmospheric aerosol components. Also discussed are some recent results of modeling the phenomena involved in Direct Containment Heating. (orig.)

  3. Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

    International Nuclear Information System (INIS)

    Durazzo, M.

    1985-01-01

    The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

  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. Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

    International Nuclear Information System (INIS)

    Samim Anghaie

    2002-01-01

    Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the core

  10. 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)

  11. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  12. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  13. 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

  14. 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).

  15. 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.

  16. 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.)

  17. 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

  18. 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)

  19. 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.)

  20. 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)

  1. 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

  2. Underground reactor containments: An option for the future?

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Kress, T.

    1997-01-01

    Changing world conditions and changing technologies suggest that serious consideration should be given to siting of nuclear power plants underground. Underground siting is not a new concept. Multiple research reactors, several weapons production reactors, and one power reactor have been built underground. What is new are the technologies and incentives that may now make underground siting a preferred option. The conditions and technologies, along with their implications, are discussed herein. Underground containments can be constructed in mined cavities or pits that are then backfilled with thick layers of rock and soil. Conventional above-ground containments resist assaults and accidents because of the strength of their construction materials and the effectiveness of their safety features that are engineered to reduce loads. However, underground containments can provide even more resistance to assaults and accidents because of the inertia of the mass of materials over the reactor. High-technology weapons or some internal accidents can cause existing strong-material containments to fail, but only very-high energy releases can move large inertial masses associated with underground containments. New methods of isolation may provide a higher confidence in isolation that is independent of operator action

  3. 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)

  4. 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)

  5. 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

  6. Intercomparison of liquid metal fast reactor seismic analysis codes. V. 3: Comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Proceedings of a final research co-ordination meeting

    International Nuclear Information System (INIS)

    1996-05-01

    This publication contains the final papers summarizing the validation of the codes on the basis of comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Refs, figs tabs

  7. Development of fast reactor metal fuels containing minor actinides

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Ogata, Takanari; Kurata, Masaki; Koyama, Tadafumi; Papaioannou, Dimitrios; Glatz, Jean-Paul; Rondinella, Vincenzo V.

    2011-01-01

    Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU. (author)

  8. Transport-diffusion coupling for Candu reactor core follow-Up

    International Nuclear Information System (INIS)

    Varin, E.; Marleau, G.; Chambon, R.

    2003-01-01

    We couple the finite reactor diffusion code DONJON and the lattice code DRAGON, called for simplicity DD, to perform reactor follow-up calculations using a history-based approach. In order to do this, a new DD module is developed. This module manages the transfer of information between standard DONJON and DRAGON data structures. Moreover, it stores in a history data structure the global and local parameters required for cell calculations as well as the isotopic composition of the various materials present in each cell of the reactor. We then implement in DD a parallel algorithm to perform history-based Candu reactor calculations. Here, we assign to each processor a specific number of fuel channels to be analyzed. The DRAGON cell calculations for each of the fuel bundles associated with the specified channels are performed on the same processor in order to minimize communication time. Only the macroscopic cross section libraries are exchanged between the processor. Since the amount of data exchanged is relatively small, we expect to obtain an ideal speed-up. The coupling is tested for the analysis of a simplified Candu reactor model with 4 x 4 channels each containing 4 bundles. A 100 full-power days core tracking sequence with 16 refueling steps is studied. Results are coherent with those obtained using more approximate approaches. Parallel speed-up is near optimal indicating that the use of this approach for more realistic reactor calculations should be pursued. (authors)

  9. 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

  10. 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)

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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

  16. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  17. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  18. Recent results on PEC reactor HCDA containment investigations

    International Nuclear Information System (INIS)

    Cenerini, R.; Palamidessi, A.; Verzelletti, G.

    1979-01-01

    The response of PEC reactor containement structures and of tank supporting arms to HCDA has been investigated by an explosive test on a refined 1:6 scaled mock-up. Experimental strains and pressures are compared with Astarte code calculations. (orig.)

  19. Dynamic analysis of reactor containment subjected to aircraft impact loading

    International Nuclear Information System (INIS)

    Li Xiaotian; He Shuyan

    2004-01-01

    In this paper, dynamic character of reactor containment subjected to aircraft impact loading is analyzed with MSC.DYTRAN program. The displacement of concrete and velocity curve of airplane is obtained. The results of the different material model are compared with empirical formula. It is concluded that reasonable result can be obtained using cap model for concrete

  20. The role of nuclear reactor containment in severe accidents

    International Nuclear Information System (INIS)

    1989-04-01

    The containment is a structural envelope which completely surrounds the nuclear reactor system and is designed to confine the radioactive releases in case of an accident. This report summarises the work of an NEA Senior Group of Experts who have studied the potential role of containment in accidents exceeding design specifications (so-called severe accidents). Some possibilities for enhancing the ability of plants to reduce the risk of significant off-site consequences by appropriate management of the acident have been examined

  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. 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. 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

  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. 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.

  8. Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

    2003-01-01

    The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

  9. Prestressed concrete nuclear reactor containment structures. Revision 3

    International Nuclear Information System (INIS)

    Reuter, H.R.; Chang-Lo, P.L.C.; Pfeifer, B.W.; Shah, G.H.; Whitcraft, J.S.

    1975-02-01

    A discussion of the techniques and procedures used for the design of prestressed concrete nuclear reactor containment structures is presented. A physical description of Bechtel designed containment structures is presented. The design bases and load combinations are given for anticipated conditions of service. Reference design documents which include industry codes, specifications, AEC Regulatory Guides, Bechtel Topical Reports and additional criteria as appropriate to containment design are listed. Stepwise procedures typically followed by Bechtel for design of containments is discussed and design examples are presented. A description of currently used analytical methods and the practical application of these methods for containment design is also presented. The principal containment construction materials are identified and codes of practice pertaining to construction procedures are listed. Preoperational structural testing procedures and post-operational surveillance programs are furnished along with results of tests on completed containment structures. (U.S.)

  10. Preparations to load, transport, receive, and store the damaged TMI-2 [Three Mile Island] reactor core

    International Nuclear Information System (INIS)

    Reno, H.W.; Schmitt, R.C.; Quinn, G.J.; Ayers, A.L. Jr.; Lilburn, B.J. Jr.; Uhl, D.L.

    1986-03-01

    The March 1979 incident at the Three Mile Island Nuclear Power Station (TMI) which damaged the core of the Unit 2 reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing, packaging, and transporting the core debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights preparations for transporting the core debris from TMI to INEL and receiving and storing that material at INEL. Issues discussed include interfacing of equipment and facilities at TMI, loading operations, transportation activities using a newly designed cask, receiving and storing operations at INEL, and criticality control during storage. Key to the transportation effort was designing, testing, fabricating, and licensing two rail casks which individually provide double containment of the damaged fuel. 27 figs

  11. 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)

  12. 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

  13. Hydrogen behavior in a large-dry pressurized water reactor containment building during a severe accident

    International Nuclear Information System (INIS)

    Hsu Wensheng; Chen Hungpei; Hung Zhenyu; Lin Huichen

    2014-01-01

    Following severe accidents in nuclear power plants, large quantities of hydrogen may be generated after core degradation. If the hydrogen is transported from the reactor vessel into the containment building, an explosion might occur, which might threaten the integrity of the building; this can ultimately cause the release of radioactive materials. During the Fukushima Daiichi nuclear accident in 2011, the primary containment structures remained intact but contaminated fragments broke off the secondary containment structures, which disrupted mitigation activities and triggered subsequent explosions. Therefore, the ability to predict the behavior of hydrogen after severe accidents may facilitate the development of effective nuclear reactor accident management procedures. The present study investigated the behavior of hydrogen in a large-dry pressurized water reactor (PWR). The amount of hydrogen produced was calculated using the Modular Accident Analysis Program. The hydrogen transport behavior and the effect of the explosion on the PWR containment building were simulated using the Flame Acceleration Simulator. The simulation results showed that the average hydrogen volume fraction is approximately 7% in the containment building and that the average temperature is 330 K. The maximum predicted pressure load after ignition is 2.55 bar, which does not endanger the structural integrity of the containment building. The results of this investigation indicate that the hydrogen mitigation system should be arranged on both the upper and lower parts of the containment building to reduce the impact of an explosion. (author)

  14. Conversion, core redesign and upgrade of the Rhode Island Atomic Energy Commission Reactor

    International Nuclear Information System (INIS)

    DiMeglio, A.F.

    1987-01-01

    The 2 MW Rhode Island Atomic Energy Commission reactor is required to convert from the use of High Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel using a standard LEU fuel plate which is thinner and contains more Uranium-235 than the current HEU plate. These differences, coupled with the fact that the conversion should be accomplished without serious degradation of reactor characteristics and capability, has resulted in core design studies and thermal hydraulic studies not only at the current 2 MW but also at the maximum power level of the reactor, 5 MW. In addition, during the course of its 23 years of operation, it has become clear that the main uses of the reactor are neutron scattering and neutron activation analysis. The requirement to convert to LEU presents an opportunity during the conversion to optimize the core for the utilization and to restudy the thermal hydraulics using modern techniques. This paper will present the preliminary conclusions of both aspects. (Author)

  15. Reactor containment depressurization and filtration equipment for use in the case of a serious accident

    International Nuclear Information System (INIS)

    L'Homme, A.

    1987-06-01

    A study was carried out under the aegis of the OECD into filtered vented containment systems which permit depressurization of the containment and filtration of the effluents released to the environment, in the event of a major accident with a pressurized water reactor (PWR) (or BWR or CANDU type reactors) involving core meltdown, with a view to minimizing the consequences. This paper describes the various systems examined which could possibly be used for this purpose. These comprised the French robust sand filtration system, the Swedish FILTRA system, the vacuum containment and discharge and emergency filtration system used by the CANDU plants of the Ontario-Hydro electricity company in Canada and the BWR pressure-suppression pounds. The positions of the various national authorities regarding incorporation of such systems into nuclear power plants, the design and technical principles underlying the systems, the procedures and criteria for their use and their advantages and disadvantages are examined [fr

  16. Code of practice for in-core instrumentation for neutron fluence rate (flux) measurements in power reactors

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    This standard applies to in-core (on-line) neutron detectors and instrumentation which is designed for safety, information or control purposes. It also applies to components in so far as these components are contained within the primary envelope of the reactor. The detector types usually used are dc ionization chambers and self-powered neutron detectors

  17. Development of a double containment concept for the European pressurized water reactor

    International Nuclear Information System (INIS)

    Costaz, J.L.; Bonhomme, N.; L'Huby, Y.; Sidaner, J.F.

    1994-01-01

    This paper addresses the development of a double containment concept for the European Pressurized Water Reactor. Specification of containment leak tightness during severe hazards resulting from core melt scenarios is part of the safety goals defined for the EPR project. These safety goals include retention of molten core, mitigation of hydrogen deflagration or explosion risks and decay heat removal. The main new containment structural design loads which have been defined, including containment pressure and temperature conditions following possible postulated-core melt events, are recalled in the paper. The feasibility of a double containment with a prestressed concrete inner containment taking into account these new design loads but based upon experience gained within the well tested concept of concrete double wall containment used in 1400 MW nuclear power plants which have already been built in France, is presented. The main characteristics of such a prestressed inner containment are described. Limits and further possible optimization for even more severe design loads (including liner option) are indicated. Experimental works including a large scale mock up are already under way. (author). 2 refs., 4 figs

  18. A remote inspection system for use inside reactor containment vessels

    International Nuclear Information System (INIS)

    Aoki, Toshihiko; Kashiwai, Jun-ichi; Yamamoto, Ikuo; Fukada, Koichi; Yamanaka, Yoshinobu.

    1985-01-01

    The harsh environment in the reactor-containment vesels of pressurized-water reactor nuclear-power plants precludes the possibility of direct circuit inspection; a remote-inspection system is essential. A robot for performing this task must not only be able to withstand the harsh conditions but must also be small and maneuverable enough to function effectively within complex and confined spaces. The article describes a monorail-type remote-inspection robot developed by Mitsubishi Electric to meet these needs, which is now under trial production and testing. (author)

  19. Application of high strength steel to nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

    1976-01-01

    Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

  20. 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

  1. 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.

  2. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  3. Exposure calculation code module for reactor core analysis: BURNER

    International Nuclear Information System (INIS)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules

  4. Proving Test on the Reliability for Reactor Containment Vessel

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.

    1988-01-01

    NUPEC (Nuclear Power Engineering Test Center) has started an eight-year project of Proving Test on the Reliability for Reactor Containment Vessel since June 1987. The objective of this project is to confirm the integrity of containment vessels under severe accident conditions. This paper shows the outline of this project. The test Items are (1) Hydrogen mixing and distribution test, (2) Hydrogen burning test, (3) Iodine trapping characteristics test, and (4) Structural behavior test. Based on the test results, computer codes are verified and as the results of analysis and evaluation by the computer codes, containment integrity is to be confirmed

  5. Hualong One's nuclear reactor core design and relative safety issues research

    Energy Technology Data Exchange (ETDEWEB)

    Yu, H., E-mail: yuhong_xing@126.com [Nuclear Power Inst. of China, Design and Research Sub-Inst., Chengdu, Sichuan (China)

    2015-07-01

    'Full text:' Hualong One, a third generation 1000MWe-class pressurized water reactor, is developed by China National Nuclear Cooperation (CNNC), based on the self-reliant technologies and experiences from China 40 years designing, construction, operation and maintenance of NPPs. In China, it has been approved to construct at Fuqing 5&6 and Fangchenggang 3&4. The Hualong One adopts advanced design features to dramatically enhance plant safety, economic efficiency and convenience of operation and maintenance. It consists of three loops with nominal thermal power output 3060 MWt and a 60-year design life. Its reactor core has 177 fuel assemblies, 18 month refueling interval (after initial cycle), and more than 15% thermal margin. It adopts low leakage loading pattern which can achieve better economy of the neutron, higher reactivity and lower radiation damage of pressure vessel. For the safety design, incorporating the feedback of Fukushima accident, the Hualong One has a combination of active and passive safety systems, a single station layout, double containment structure, and comprehensive implementation of defence-in-depth design principles. The new design features has been successfully evaluated to ensure that they enhance the performance and safety of Hualong One. Several experimental activates have been conducted, such as cavity injection and cooling system testing, passive containment heat removal system testing, and passive residual heat removal system of secondary side testing. The future improvements of Hualong reactor will focus on better economic core design and more reliable safety system. (author)

  6. 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

  7. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  8. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

    International Nuclear Information System (INIS)

    Kim, Tae H.; Yang, Chae Y.

    2013-01-01

    The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

  11. A model to predict moisture conditions in concrete reactor containments

    International Nuclear Information System (INIS)

    Ahs, M.; Nilsson, L.O.; Poyet, S.; L'Hostis, V.

    2015-01-01

    Moisture has an impact in many of the degradation mechanisms that appear in the structures of a nuclear power plant. Moisture conditions in a reactor containment wall have been simulated by using a hygro-thermal model of drying concrete. Methods to estimate the temperature dependency of the sorption isotherms and moisture transport properties is suggested and applied in the model. This temperature dependency is included as there is a temperature gradient present through the containment wall. The hygro-thermal model was applied on a full scale 3D model of a real reactor containment building and the concrete relative humidity has been computed at 4 different moments: 1, 10, 20 and 30 years. The results show that the major part of the concrete is not dried at all even after 30 years of operation. It is also clear that the temperature distribution inside the whole concrete volume is affected by the variable boundary conditions. It was concluded that the suggested hygro-thermal model was appropriate to use as a method to estimate the existing conditions in a PWR reactor containment wall

  12. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    Energy Technology Data Exchange (ETDEWEB)

    Prabhudharwadkar, Deoras M. [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Iyer, Kannan N., E-mail: kiyer@iitb.ac.i [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Mohan, Nalini; Bajaj, Satinder S. [Nuclear Power Corporation of India Ltd., Mumbai (India); Markandeya, Suhas G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India)

    2011-03-15

    Research highlights: This work addresses hydrogen dispersion in commercial nuclear reactor containment. The numerical tool used for simulation is first benchmarked with experimental data. Parametric results are then carried out for different release configurations. Results lead to the conclusion that the dispersal is buoyancy dominated. Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is quite poor and most

  13. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    International Nuclear Information System (INIS)

    Prabhudharwadkar, Deoras M.; Iyer, Kannan N.; Mohan, Nalini; Bajaj, Satinder S.; Markandeya, Suhas G.

    2011-01-01

    Research highlights: → This work addresses hydrogen dispersion in commercial nuclear reactor containment. → The numerical tool used for simulation is first benchmarked with experimental data. → Parametric results are then carried out for different release configurations. → Results lead to the conclusion that the dispersal is buoyancy dominated. → Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is

  14. 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)

  15. 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)

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. Actions to reduce radioactive emissions: prevention of containment failure by flooding Containment and Reactor Cavity; Acciones para la reduccion de emisiones radiactivas: prevencion del fallo de la Contencion mediante la inundacion de la Contencion y de la Cavidad del Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fornos Herrando, J.

    2013-07-01

    The reactor cavity of Asco and Vandellos II is dry type, thus a severe accident leading to vessel failure might potentially end up resulting in the loss of containment integrity, depending on the viability to cool the molten core. Therefore, significant radioactive emissions could be released to outside. In the framework of Fukushima Stress Tests, ANAV has analyzed the convenience of carrying out different actions to prevent failure of the containment integrity in order to reduce radioactive emissions. The aim of this paper is to present and describe the main phenomenological aspects associated with two of these actions: containment flooding and reactor cavity flooding.

  4. 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

  5. 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.

  6. 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

  7. 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

  8. TMI in perspective: reactor containment stands up, difficult decisions remain

    International Nuclear Information System (INIS)

    Corey, G.R.

    1979-01-01

    Commonwealth Edison Co. is increasing its commitment to nuclear energy after reviewing the performance of the Three Mile Island reactor containment systems. Both the reactor vessel and the secondary containment remained intact and no radiation was reported in the soil or water. The public discussion of energy options which followed the accident will benefit both the public and technical community even if there is a temporary slowdown in nuclear power development. The realities of energy supplies have become evident; i.e., that nuclear and coal are the only available options for the short-term. The discussion should also lead to better personnel training, regulatory reforms, risk-sharing insurance, and international standards. The public hysteria triggered by the accident stemmed partly from the combination of unfortunate incidents and the media coverage, which led to hasty conclusions

  9. Contained fissionly vaporized imploded fission explosive breeder reactor

    International Nuclear Information System (INIS)

    Marwick, E.F.

    1978-01-01

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

  10. Structural design and dynamic analysis of underground nuclear reactor containments

    International Nuclear Information System (INIS)

    Kierans, T.W.; Reddy, D.V.; Heale, D.G.

    1975-01-01

    Present actual experience in the structural design of undeground containments is limited to only four rather small reactors all located in Europe. Thus proposals for future underground reactors depend on the transposition of applicable design specifications, constraints and criteria from existing surface nuclear power plants to underground, and the use of many years of experience in the structural design of large underground cavities and cavity complexes for other purposes such as mining, hydropower stations etc. An application of such considerations in a recent input for the Underground Containment sub-section of the Seismic Task Group Report to the ASCE Committee for Nuclear Structures and Materials is presented as follows: underground concept considerations, siting criteria and structural selection, structural types, analytical and semi-analytical approaches, design and other miscellaneous considerations

  11. Uncertainties assessment for safety margins evaluation in MTR reactors core thermal-hydraulic design

    International Nuclear Information System (INIS)

    Gimenez, M.; Schlamp, M.; Vertullo, A.

    2002-01-01

    This report contains a bibliographic review and a critical analysis of different methodologies used for uncertainty evaluation in research reactors core safety related parameters. Different parameters where uncertainties are considered are also presented and discussed, as well as their intrinsic nature regarding the way their uncertainty combination must be done. Finally a combined statistical method with direct propagation of uncertainties and a set of basic parameters as wall and DNB temperatures, CHF, PRD and their respective ratios where uncertainties should be considered is proposed. (author)

  12. Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

    International Nuclear Information System (INIS)

    Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

    1993-12-01

    Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

  13. 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

  14. Study on confinement function of reactor containment during late phase severe accident

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    During a severe accident reactor containment integrity is maintained by accident management. However, gas leakage from containment is inevitable after the severe accident. A large amount of hydrogen and rare gases are produced due to core damage or melting. These non-condensable gases cause the containment pressure much higher than atmospheric pressure even after residual heat removal system recovery especially for BWR with smaller containment volume. Besides, iodine confined in water pool is re-evaporated under radiation field. The present study consists of realistic evaluation of fission products source term inside containment, quantitative evaluation of iodine re-evaporation effect and the experimental study of hydrogen treatment in BWR using ammonia production method by catalyst. Activities in fiscal year 2012 are that modification of MELCOR fission product chemical model was done and verified by experimental data, and that effects of CsI on ammonia production rate for Ru catalyst were conducted. (author)

  15. 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)

  16. 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)

  17. Japanese contributions to containment structure, assembly and maintenance and reactor building for ITER

    International Nuclear Information System (INIS)

    Shibanuma, Kiyoshi; Honda, Tsutomu; Kanamori, Naokazu

    1991-06-01

    Joint design work on Conceptual Design Activity of International Thermonuclear Experimental Reactor (ITER) with four parties, Japan, the United States, the Soviet Union and the European Community began in April 1988 and was successfully completed in December 1990. In Japan, the home team was established in wide range of collaboration between JAERI and national institute, universities and heavy industries. The Fusion Experimental Reactor (FER) Team at JAERI is assigned as a core of the Japanese home team to support the joint Team activity and mainly conducted the design and R and D in the area of containment structure, remote handling and plant system. This report mainly describes the Japanese contribution on the ITER containment structure, remote handling and reactor building design. Main areas of contributions are vacuum vessel, attaching locks, electromagnetic analysis, cryostat, port and service line layout for containment structure, in-vessel handling equipment design and analysis, blanket handling equipment design and related short term R and D for assembly and maintenance, and finally reactor building design and analysis based on the equipment and service line layout and components flow during assembly and maintenance. (author)

  18. Evaluation of a hydrogen sensor for nuclear reactor containment monitoring

    International Nuclear Information System (INIS)

    Hoffheins, B.S.; McKnight, T.E.; Lauf, R.J.; Smith, R.R.; James, R.E.

    1997-01-01

    Measurement of hydrogen concentration in containment atmospheres in nuclear plants is a key safety capability. Current technologies require extensive sampling systems and subsequent maintenance and calibration costs can be very expensive. A new hydrogen sensor has been developed that is small and potentially inexpensive to install and maintain. Its size and low power requirement make it suitable in distributed systems for pinpointing hydrogen buildup. This paper will address the first phase of a testing program conducted to evaluate this sensor for operation in reactor containments

  19. Evaluation of materials for retention of sodium and core debris in reactor systems. Annual progress report, September 1977-December 1978

    International Nuclear Information System (INIS)

    Swanson, D.G.; Zehms, E.H.; McClelland, J.D.; Meyer, R.A.; van Paassen, H.L.L.

    1978-12-01

    This report considers some of the consequences of a hypothetical core disruptive accident in a nuclear reactor. The interactions expected between molten core debris, liquid sodium, and materials that might be employed in an ex-vessel sacrificial-bed or in the reactor building are discussed. Experimental work performed for NRC by Sandia Laboratories and Hanford Engineering Development Laboratory on the interactions between liquid sodium and basalt concrete is reviewed. Studies of molten steel interactions with concrete at Sandia Laboratories and molten UO 2 interactions with concrete at The Aerospace Corporation are also discussed. The potential of MgO for use in core containment is discussed and refractory materials other than MgO are reviewed. Finally, results from earlier experiments with molten core debris and various materials performed at The Aerospace Corporation are presented

  20. Synthesis of Ni-SiO2/silicalite-1 core-shell micromembrane reactors and their reaction/diffusion performance

    KAUST Repository

    Khan, Easir A.

    2010-12-15

    Core-shell micromembrane reactors are a novel class of materials where a catalyst and a shape-selective membrane are synergistically housed in a single particle. In this work, we report the synthesis of micrometer -sized core-shell particles containing a catalyst core and a thin permselective zeolite shell and their application as a micromembrane reactor for the selective hydrogenation of the 1-hexene and 3,3-dimethyl-1-butene isomers. The bare catalyst, which is made from porous silica loaded with catalytically active nickel, showed no reactant selectivity between hexene isomers, but the core-shell particles showed high selectivities up to 300 for a 1-hexene conversion of 90%. © 2010 American Chemical Society.